[Federal Register Volume 59, Number 32 (Wednesday, February 16, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-3465]


[[Page Unknown]]

[Federal Register: February 16, 1994]


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NUCLEAR REGULATORY COMMISSION

 

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 22, 1994, through February 4, 1994. 
The last biweekly notice was published on February 2, 1994 (59 FR 
4933).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 P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies 
of written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By March 18, 1994, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendment request: January 24, 1994
    Description of amendment request: The proposed amendment would 
implement Line Item 5.9 of NRC Generic Letter 93-05, ``Line Item 
Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation,'' which recommends 
licensees consider deleting the requirements to perform response time 
testing for selected instrumentation in the isolation system where the 
required time corresponds to the diesel start time.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    LaSalle has evaluated the proposed Technical Specification 
Amendment. Based upon the criteria for defining a Significant 
Hazards Consideration established in 10 CFR 50.92(c), operation of 
LaSalle County Station in accordance with the proposed amendment 
will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposal seeks to eliminate response time testing 
requirements for selected instrumentation in the isolation system. 
The proposal does not introduce changes in the response times 
themselves. The probability and consequences of an accident 
previously evaluated are not increased because accepted licensing 
criteria are maintained. The requirements for channel checks, 
functional tests, calibrations, and logic system functional tests 
are not altered by this proposal. The ability to detect degrading 
trends of response times is available via the above Technical 
Specification required tests. Therefore, the response times of these 
systems will be maintained within the acceptance limits assumed in 
plant safety analyses and required for successful mitigation of an 
initiating event because of the continued Technical Specification 
testing.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposal does not change component or system interactions. 
Accident analyses assume a loss of AC power which is restored by 
startup of emergency diesel generators. The 13 second interval 
associated with the restoration of AC power, which establishes the 
response time for the isolation functions, is maintained. The 
starting, sequencing, and loading functions associated with the 
diesel generators is not affected by the proposed change. The 
response times include the instrument response times, which are 
typically measured in fractions of a second, and the response times 
of the actuation logic circuits, which are typically less than a 
second. These times are small in comparison to the diesel generator 
start time (13 seconds). The ability of the isolation system to 
perform its intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in plant 
safety analyses is not altered by the proposed change.
    3) Involve a significant reduction in a margin of safety 
because:
    The proposal does not involve the relaxation of any criteria 
identified in the SAR or reduce any of the requirements of Technical 
Specifications. The proposed revision does not affect licensing 
acceptance limits associated with accidents. With the exception of 
MSIVs, the safety analyses do not address individual sensor response 
times or the response times of the logic systems to which the 
sensors are connected. These analyses conservatively establish the 
margin of safety. Deleting the requirement to perform unnecessary 
response time testing does not affect the results of accident and 
transient analyses. Plant and system response to an initiating event 
will remain in compliance within the assumptions of safety analyses.
    The proposed change does not increase the probability or 
consequences of an accident, and there is no impact on equipment 
important to safety or systems, structures or components. There is 
no associated change to the type, amount, or control of radioactive 
effluents, nor is there an associated increase in individual or 
cumulative occupational radiation exposure. There is no effect upon 
the capabilities of the associated systems to perform their intended 
functions within the allowed response times assumed in safety 
analyses. Therefore, the margin of safety is preserved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Ogelsby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of amendment request: January 17, 1994
    Description of amendment request: Connecticut Yankee Atomic Power 
Company (CYAPCO) proposes to remove Technical Specification 3/4.4.12, 
``Failed Fuel Rods'' and its associated BASES Section 3/4.4.12.
    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 (SHC), which is presented below:
    The proposed changes do not involve an SHC consideration because 
the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A review of the accidents detailed in the Updated Final Safety 
Analysis Report, Chapter 15, was undertaken to determine if they 
were impacted by the proposed change. The review indicated that the 
previously evaluated accidents were not impacted by the proposed 
license amendment.
    All fuel design and performance criteria are the same for Cycle 
18 as in previous cycles. All criteria will continue to be met and 
no new single-failure mechanisms will be created. This change does 
not involve any alterations to plant equipment or procedures which 
would affect any operational modes or accident assumptions. This 
proposed license amendment does delete a technical specification 
that is no longer considered necessary. This deletion is prompted by 
the replacement of stainless steel clad fuel with zircaloy clad 
fuel. The zircaloy clad fuel, if it experiences damage, will release 
iodine into the primary system. Any iodine released is covered 
within the guidelines specified in the existing Technical 
Specification 3/4.4.8, ``Specific Activity.'' This specification 
will ensure that operation does not continue with radiochemistry 
values that exceed those assumed in our accident assumptions. The 
existing Technical Specification of specific activity along with the 
zircaloy clad fuel will ensure that a significant increase in the 
probability or consequences of an accident previously evaluated is 
not present.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The possibility of an accident or malfunction of a different 
type than any evaluated previously in the UFSAR [Updated Final 
Safety Analysis Report] is not created. Since there are no changes 
in the way the plant is operated, the potential for an unanalyzed 
accident is not created. No new failure modes are introduced.
    The presence of defective fuel rods and the resultant iodine 
release would only affect potential offsite doses. This proposed 
license amendment does not increase the radiochemistry limits, but 
does revert the technical specifications back to the standard 
methodology and limitations that were unable to be used because of 
the stainless steel clad fuel. These new limitations will continue 
to ensure that doses remain within the limits prescribed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not have any adverse impact on the 
protective boundaries. The margin of safety, as defined in the basis 
for any technical specification, is not reduced. The proposed 
changes do not adversely impact any of the safety systems, nor do 
they increase the number of challenges to the safety systems.
    The limit of 160 defective rods was chosen to be consistent with 
initial conditions assumed for the radiological design basis. The 
elimination of this specification is acceptable since the basis for 
the initial condition can be supported by the use of zircaloy clad 
fuel as opposed to the unique stainless steel clad. If future fuel 
defects are debris induced, the dose equivalent iodine will be 
within expected radiochemistry values and the resulting doses will 
be bounded. Therefore, there is no reduction in the margin of safety 
as defined in the basis of any technical specification with the 
deletion of the defective fuel rod technical specification.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: December 6, 1993
    Description of amendment request: The proposed amendment request 
would revise the Technical Specifications (TSs) to provide a temporary 
one-time revision to the Definition Section of the TS. Specifically, a 
footnote is added in the Definition Section of the TS which is 
applicable to TS 1.2.1, ``Cold Shutdown Condition,'' changing Tavq 
less than or equal to 200 deg.F to less than or equal to 250 deg.F and 
TS 1.2.2, ``Hot Shutdown Condition,'' changing Tavq greater than 
200 deg.F to greater than 250 deg.F. The footnote further states that 
the change is for the one time, fuel out, chemical decontamination 
program. This program is currently scheduled for the upcoming 1995 
refueling outage of the Indian Point Nuclear Generating Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    Approval of the proposed one time change to the Technical 
Specification definition of cold shutdown for purposes of performing 
the full RCS [reactor coolant system] chemical decontamination 
without fuel in the reactor would provide relief from unnecessary 
technical specification action statements that are based on fuel in 
the reactor. Credible accidents with significant consequences are 
practically eliminated with the removal of the reactor fuel during 
the performance of the FSD [full reactor coolant system chemical 
decontamination]. In addition, specific actions would be taken in 
accordance with the requirements of the NRC approved WCAP-12932-A 
Rev. 2 to ensure that RCS and affected interfacing systems integrity 
are preserved. Thus, system capability within established accident 
scenarios would not be compromised. The proposed amendment would 
therefore not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    As noted above, the proposed amendment seeks to eliminate 
unnecessary Technical Specification action requirements during the 
performance of full RCS chemical decontamination. These actions are 
unnecessary because there will be no fuel in the reactor and the RCS 
and other affected systems will be operated under conditions well 
within their design capability during the implementation of this 
process. In addition, the FSD effort will be conducted in accordance 
with the requirement(s) of the NRC approved Westinghouse topical 
report WCAP-12932-A Rev. 2. Accidents involving failures of the 
decontamination process system will not exceed the bounding 
conditions for any previously established accidents involving 
failure of a radwaste system. Accordingly, the possibility of a new 
or different kind of accident from any previously analyzed will not 
be created.
    3. There has been no reduction in the margin of safety.
    The proposed amendment provides relief from technical 
specification actions in the performance of the FSD which become 
unnecessary when there is no fuel in the reactor. The change will 
not adversely impact any Technical Specification required systems, 
structures or components. The design capability of systems, 
structures or components impacted will not be reduced. Consequently, 
no significant reduction in the margin of safety for any system, 
structure, or component is involved.
    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 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: January 18, 1994
    Description of amendment request: The proposed amendments remove 
the tables of containment penetration conductor overcurrent protective 
devices from the Technical Specifications (TS) in accordance with the 
guidance contained in Generic Letter 91-08, ``Removal of Component 
Lists from Technical Specifications.'' The tables would be relocated to 
Chapter 16 of the Catawba Final Safety Analysis Report (Selected 
Licensee Commitments Manual). In addition, the licensee proposes the 
removal of an obsolete footnote to TS 4.8.4. The footnote, which made 
TS 4.8.4.a initially effective following the first refueling outage of 
Unit 1, is no longer needed.
    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. Relocating the component lists of containment penetration 
conductor overcurrent protective devices from the technical 
specifications to the [Selected Licensee Commitments] SLC Manual 
(with all attendant required technical specification changes as 
described previously and also including removal of the above 
described obsolete footnote) has no impact upon either the 
probability or consequences of any accident. No plant equipment is 
affected by the proposed change. No equipment is being added or 
deleted from the lists; only the source document for the lists is 
being changed. Any future changes to the lists (i.e., changes to the 
plant) will be subject to the provisions of 10CFR50.59 and also 
subject to the change control provisions of Chapter 6 of Catawba's 
Technical Specifications.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No accident causal mechanisms are affected by the 
proposed change, as no change to the plant is being proposed. In 
addition, no change to the manner in which the plant is operated is 
being made. Finally, no changes to plant procedures are being made 
which would affect any accident causal mechanisms.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. The proposed change has no impact 
upon any safety margin. The proposed change is consistent with the 
guidance provided in Generic Letter 91-08 and the control provisions 
utilized as a result of relocating the subject component lists are 
at least as stringent as those set forth in the generic letter.
    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: Loren R. Plisco, Acting

Duke Power Company, Docket No. 50-413, Catawba Nuclear Station, 
Unit No. 1, York County, South Carolina

    Date of amendment request: January 10, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 2.0 and 3/4.2 which currently requires 
the determination of the reactor coolant system flow rate by precision 
heat balance measurement at least once per 18 months. Date of 
publication of individual notice in Federal Register: January 26, 1994 
(59 FR 3743)
    Expiration date of individual notice: February 25, 1994
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of amendment request: January 10, 1994
    Description of amendment request: The amendments would change the 
method of measuring the reactor coolant system flow rate (Technical 
Specifications 2.0 and 3/4.2) during the 18-month surveillance for 
McGuire, Units 1 and 2.
    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) This amendment will not significantly increase the probability 
or consequence of any accident previously evaluated.

    No component modification, system realignment, or change in 
operating procedure will occur which could affect the probability of 
any accident or transient. The change in method of flow measurement 
will not change the probability of actuation of any Engineered 
Safeguard Feature or other device. The actual flow rate will not 
change. The consequences of previously-analyzed accidents will not 
change as a result of the new method of flow measurement.

(2) This amendment will not create the possibility of any new or 
different accidents not previously evaluated.

    No component modification or system realignment will occur which 
could create the possibility of a new event not previously 
considered. The elbow taps are already in place, and are used to 
monitor flow for the Reactor Protection System. They will not 
initiate any new events.

(3) This amendment will not involve a significant reduction in a 
margin of safety.

    As described in [the licensee's application], the change in 
method of RCS flow measurement will provide a more accurate 
indication of the flow. The actual flow rate will not be affected. 
The revised setpoints for low reactor coolant flow are driven by 
changes to statistical allowances and do not represent substantive, 
or less conservative, changes. There is no 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: Astkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Loren R. Plisco, Acting

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

    Date of amendment request: December 23, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) for the following four items in 
accordance with the guidance in Generic Letter (GL) 93-05 ``Line Item 
Technical Specifications Improvements To Reduce Surveillance 
Requirements For Testing During Power Operation''.1) GL Item 5.14 
Radiation Monitors will change the channel functional test from monthly 
to quarterly.2) GL Item 6.1 Reactor Coolant System (RCS) Isolation 
Valves will increase the time from 72 hours to 7 days for remaining in 
cold shutdown without leak testing the RCS isolation valves.3) GL Item 
6.6 Pressurizer Heaters will change the verification of capacity from 
at least once per 92 days to each refueling outage and will change the 
demonstration of the emergency power supply from at least once per 18 
months to at each refueling outage.4) GL Item 9.1 Auxiliary Feedwater 
Pump and System Testing will change the frequency of these pumps from 
once per 31 days on a staggered basis to quarterly on a staggered 
bases.
    All of the above are compatible with Waterford 3 plant operating 
experience and are consistent with NUREG-1366, ``Improvement To 
Technical Specification Surveillance Requirements,'' December 1992 and 
the licensing basis for Waterford 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change to increase the radiation monitoring 
instrumentation channel functional test from monthly to quarterly 
will have no effect on design basis accidents. The findings in 
NUREG-1366 determined that this change will increase the 
availability of radiation monitors.
    The proposed change to increase the 72 hour time for remaining 
in cold shutdown without leak testing the RCS isolation valves to 7 
days will not affect any design basis accidents. NUREG-1366 findings 
have determined that extending this interval does not significantly 
alter the associated risk. In addition, the current requirement has 
a potential for causing problems resulting from a hurried recovery.
    The proposed change to the pressurizer heater capacity test 
interval from quarterly to each refueling interval will have no 
affect on any design basis accidents. The TS requires at least 2 
groups of pressurizer heaters each having a nominal capacity of 150 
kW. Waterford 3 has 8 groups of pressurizer heaters; two 
proportional groups of 150 kW each, and 6 backup groups of 200 kW 
each. An evaluation of past operating experience has shown the 
availability of at least 6 groups of pressurizer heaters with a 
minimum of 150 kW each.
    The proposed change to extend the testing interval for the EFW 
[emergency feedwater] pumps will have no affect on any design basis 
accidents. The pumps will continue to be tested quarterly to the 
same standards applied to safety related pumps as defined by the 
ASME [American Society of Mechanical Engineers] Section XI Code. 
Satisfactory completion of testing in accordance with the Code is 
accepted as verification that safety related pumps will be available 
to perform their intended function.
    The proposed changes identified above are supported by the 
findings identified in NUREG-1366 and consistent with the guidance 
provided in Generic Letter 93-05. These line-item improvements are 
intended to improve plant safety, decrease equipment degradation, 
and remove unnecessary burden on personnel resources by reducing the 
amount of testing that the TS require during power operation. 
Therefore, the proposed changes identified above will not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    The changes identified above only affect the frequency of 
surveillance testing. There are no changes that will alter operation 
of the plant or the manner in which it is operated. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes identified herein extend testing frequency 
in an effort to improve plant reliability and safety. The proposed 
changes are consistent with the findings in NUREG-1366, guidance in 
Generic Letter 93-05 and plant operating experience. As such, the 
proposed changes will preserve the established margin of safety for 
the affected specifications. Therefore, the proposed changes will 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N. S. Reynolds, Esq., Winston & Strawn 1400 
L Street NW., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 19, 1993
    Description of amendment request: The proposed change would 
relocate the requirements of Technical Specification 3/4.3.4, Turbine 
Overspeed Protection, to Section 16.3 of the Vogtle Electric Generating 
Plant, Units 1 and 2, Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change involves the relocation of the TS 
[Technical Specification] requirements for the turbine overspeed 
protection system to the VEGP [Vogtle Electric Generating Plant] 
FSAR [Final Safety Analysis Report]. The requirements that will 
reside in the FSAR will continue to ensure that the probability of 
turbine missile generation is maintained below NRC limits as defined 
in NUREG-1048, Appendix U. Since the turbine overspeed protection 
system will remain capable of protecting the turbine from excessive 
overspeed, the proposed change will have no effect on the 
consequences of an accident previously evaluated.
    2. The proposed change will not create the possibility of a new 
or different kind of accident than any previously evaluated. The 
proposed change does not involve any change to the configuration or 
method of operation of any plant equipment, and no new failure modes 
have been defined for any plant system or component. In addition, no 
new limiting failures have been identified as a result of the 
proposed change. The requirements for the turbine overspeed 
protection system that will reside in the FSAR will ensure that the 
system remains capable of protecting the turbine from excessive 
overspeed. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident than any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety. The proposed change would allow the 
requirements for the turbine overspeed protection system to be 
relocated to the FSAR on the basis that the turbine overspeed 
protection system does not meet the criteria of the NRC Final Policy 
Statement on Technical Specifications Improvements for Nuclear 
Reactors. The requirements that will reside in the FSAR for the 
turbine overspeed protection system will ensure that the system 
remains capable of protecting the turbine from excessive overspeed. 
Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Loren R. Plisco, Acting

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket No. 50-499 South Texas Project, Unit 2, Matagorda County, 
Texas

    Date of amendment request: January 25, 1994
    Description of amendment request: The licensee proposes to make a 
one-time change to the technical specifications to add new Technical 
Specifications 3/4.10.6 and 3/4.10.7 to the Special Test Exemptions 
section. The new TS would allow the restart of Unit 2 with expired 
calibrations on the core exit thermocouples (CET) and the reactor 
coolant system (RCS) resistance temperature detectors (RTD). This 
amendment will also add a new Technical Specification to allow the 
ascension to 75 percent rated thermal power with an expired precision 
heat balance reactor coolant flow measurement.
    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 a previously evaluated 
accident.
    The proposed change will allow the restart of [STP] Unit 2 with 
Core Exit Thermocouples and Reactor Coolant System Resistance 
Temperature Detectors technically inoperable due to expired 
calibrations. The calibrations of these instruments can only be 
completed when the Unit reaches Normal Operating Pressure and Normal 
Operating Temperature in Mode 3. Once the calibrations of these 
instruments are completed, this one time change will expire and all 
of the existing applicable Limiting Conditions for Operations will 
become effective immediately. Since industry and South Texas Project 
Electric Generating Station experience has shown that the failure 
mechanism for these types of instrument is complete failure as 
opposed to a gradual drift, and there will be calibration points to 
compare RTD readings to actual RCS temperature as the RCS 
temperature increases, it is reasonable to expect these CETs/RTDs 
will function as they did before their calibrations expired. For 
this reason, all applicable functions, including COMS, Thot , 
Tcold, and Tavg are expected to operate normally. Because 
normal operation of the instruments is expected and the only reason 
for the instruments being declared inoperable is their expired 
calibrations, this change does not involve a significant increase in 
the probability or consequence of an accident previously evaluated.
    The proposed change will also allow the restart of Unit 2 with 
the precision heat balance RCS flow measurement surveillance 
expired. This surveillance is used to confirm the values indicated 
by the RCS flow meters. These instruments are calibrated every 18 
months and the RCS flow meters will be checked every 12 hours to 
ensure adequate flow prior to the completion of the precision heat 
balance RCS flow measurement. Since this surveillance is only used 
to confirm the reading of calibrated instruments and does not 
involve any changes to the design or function of the instruments, 
this 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 operations of Unit 2 with the CETs and RCS RTDs technically 
inoperable due to expired calibrations, until these calibrations can 
be completed in Mode 3, does not affect the design bases of the CETs 
and RCS RTDs or any of the accident evaluations involving these 
instruments. Since industry and South Texas Project Electric 
Generating Station experience indicates that the failure mechanism 
for these types of instruments is not a gradual drift but complete 
failure, the reasonable expectation is the CETs/RTDs will function 
as they did prior to their calibrations expiring.
    Additionally, the operation of Unit 2 with the precision heat 
balance RCS flow measurement surveillance expired does not affect 
the design bases of the RCS flow meters or any of the accident 
evaluations involving these instruments. This surveillance is used 
to confirm the values indicated by the RCS flow meters. These 
instruments are calibrated every 18 months and the RCS flow meters 
will be checked every 12 hours to ensure adequate flow prior to the 
completion of the precision heat balance RCS flow measurement.
    Because normal operation of all of these instruments is 
expected, these changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in the margin of safety.
    The RCS RTDs are auctioneered to prevent a failed high or low 
instrument from adversely influencing the safety of the plant. This 
feature is still operable and will, along with normal operator 
activities, provide assurance that the margin of safety is not 
reduced by this change. In addition, the change does not affect the 
design bases, accident analysis, reliability or capability of the 
CETs/RTDs to perform their intended safety functions. The RCS flow 
meters will be checked every 12 hours to ensure adequate flow prior 
to the completion of the precision heat balance RCS flow 
measurement.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendment involves no significant hazards 
consideration.Local Public Document Location: Wharton County Junior 
College, J.M. Hodges Learning Center, 911 Boling Highway, Wharton, 
Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW, Washington, DC 20036
    NRC Project Director: Suzanne C. Black

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

    Date of amendment request: September 28, 1993
    Description of amendment request: The proposed amendment would 
revise the Cooper Nuclear Station (CNS) Technical Specifications to 
modify the licensee's organizational structure by removing the 
positions of ``Site Manager'' and ``Senior Manager of Operations.'' The 
functions presently given in CNS Technical Specifications for the Site 
Manager position will be assumed by the Vice President - Nuclear.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Evaluation
    The proposed change removing the positions of Site Manager and 
Senior Manager of Operations from the Technical Specifications is 
administrative in nature. The functions and responsibilities of the 
previous position of Site Manager presently given in the plant 
Technical Specifications will be performed by the Vice President - 
Nuclear. Additionally, with the reorganization, the Senior Manager 
of Operations position is eliminated and therefore, this position is 
also being removed. The provision in the Technical Specifications 
for automatic shifting of Plant Manager responsibilities to the 
Senior Manager of Operations has also been removed. The shifting of 
Plant Manager responsibilities (in writing) to one of the Managers 
at CNS who is qualified for this position remains in the Technical 
Specifications. The position removals and responsibility transfers 
in the organization do not affect plant design or operation, nor do 
they affect the way any systems, structures, or components are 
operated or maintained. The individual filling the position ``Vice 
President - Nuclear'' is qualified to perform the assigned tasks and 
responsibilities. Restructuring of the sentence in specification 
6.2.B.6, is purely an administrative change. Also, this proposed 
change does 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 consequences previously 
evaluated are not adversely affected by the proposed change. 
Therefore, it can be concluded that the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed License Amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Evaluation
    The proposed Technical Specification revision removes all 
references to the position title of the Site Manager. The 
responsibilities of this position presently given in the Technical 
Specifications are being incorporated and performed by the position 
``Vice President - Nuclear.'' Additionally, with the reorganization, 
the Senior Manager of Operations position is eliminated and 
therefore, this position is also being removed. The shifting of 
Plant Manager responsibilities (in writing) to one of the Managers 
at CNS who is qualified for this position remains in the Technical 
Specifications. All given management activities will continue to be 
performed by qualified individuals. Restructuring of the sentence in 
specification 6.2.B.6 is purely an administrative change. This 
change does not affect the design or operation of any system, 
structure, or component in the plant, and is considered to be an 
administrative change. Accordingly, no new failure modes have been 
defined for any plant system or component important to safety, nor 
has any new limiting failure been identified as a result of the 
proposed change. Therefore, this proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Evaluation
    This proposed amendment involves a change to the Administrative 
Controls Section of the CNS Technical Specifications; specifically, 
removal of two positions referenced in the organizational structure. 
The Site Manager position is being deleted and the responsibilities 
of this position listed in the Technical Specifications are being 
performed by the Vice President - Nuclear. Additionally, with the 
reorganization, the Senior Manager of Operations position is 
eliminated and therefore, this position and responsibilities are 
also being removed. The shifting of Plant Manager responsibilities 
(in writing) to one of the Managers at CNS who is qualified for this 
position remains in the Technical Specifications. All given 
management activities, as described in the Technical Specifications, 
will continue to be performed by qualified individuals. 
Restructuring of the sentence in specification 6.2.B.6, is purely an 
administrative change. The proposed change does not adversely impact 
the plant's ability to meet applicable regulatory requirements. The 
proposed change does not alter any means of plant operation, nor 
does the proposed change involve any physical alterations to the 
plant and does not affect any plant safety parameters or setpoints. 
Therefore, this proposed 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: Asuburn Public Library, 118 
15th Street, Auburn, Nebraska 68305
    Attorney for licensee: Mr. G. D. Watson, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 10, 1993
    Description of amendment request: The proposed amendment would 
revise the Cooper Nuclear Station (CNS) Technical Specifications 
Sections 3/4.21 ``Environmental/Radiological Effluents,'' and 6.5, 
``Station Reporting Requirements,'' to change the frequency of the 
reporting period of the ``Semiannual Radioactive Materials Release 
Report'' from semiannual to annual and to extend the reporting 
frequency of the Annual Design Change Report from an annual submittal 
to annually or along with the Updated Safety Analysis Report (USAR) 
updates required by 10 CFR 50.71(e). These proposed changes are 
intended to make the CNS Technical Specifications consistent with the 
current provisions of 10 CFR 50.36(a) and 10 CFR 50.59(b), 
respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Evaluation
    The proposed changes are administrative in nature and makes the 
Cooper Nuclear Station (CNS) Technical Specifications (T/S) 
consistent with amended regulations of 10CFR50.36(a), and 10CFR 
50.59(b) by reducing the submittal frequency of certain reports to 
the NRC. The proposed revisions do not involve any change to plant 
design, plant operation, or configuration of any plant equipment 
that is used to mitigate the consequences of an accident previously 
evaluated. Also, 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. As administrative changes, all 
defined terms on the affected pages have been capitalized. 
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. Does the proposed License Amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Evaluation
    The proposed changes are administrative in nature and makes the 
CNS T/S consistent with amended regulations of 10CFR50.36(a), and 
10CFR50.59(b) by reducing the submittal frequency of certain reports 
to the NRC. The proposed revisions do not involve any change to 
plant design, plant operation, or configuration of any plant 
equipment that is used to mitigate the consequences of an accident 
previously evaluated. Accordingly, no new failure modes have been 
created for any plant system or component important to safety nor 
has any new limiting failure been identified as a result of the 
proposed changes. Also, there will be no change in the types or 
increase in the amount of effluents released offsite. As 
administrative changes, all defined terms on the affected pages have 
been capitalized. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change create a significant reduction in 
the margin of safety?
    Evaluation
    The proposed changes are administrative in nature and do not 
adversely impact the plant's ability to meet applicable regulatory 
requirements related to liquid or gaseous effluents, and solid waste 
releases. The proposed changes do not alter any administrative 
controls over radioactive effluents, nor do the proposed changes 
involve any physical alterations to the plant with respect to 
radioactive effluents. These changes do not affect the meaning, 
application, and function of the T/S requirements. The proposed 
change will reduce the administrative burden of NRC reporting 
without reducing the protection for public health and safety. As 
administrative changes, all defined terms on the affected pages have 
been capitalized. Therefore, the proposed 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: Asuburn Public Library, 118 
15th Street, Auburn, Nebraska 68305
    Attorney for licensee: Mr. G. D. Watson, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499
    NRC Project Director: William D. Beckner

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

    Date of amendment request: January 6, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Tables 3.2.7, 3.6.2a, 4.6.2a, 
3.6.2b and 4.6.2b to delete the main steam line isolation and automatic 
reactor shutdown (reactor scram) functions of the Main Steam Line 
Radiation Monitor. Conforming changes would also be made to the Bases 
of these TSs and to the Bases for TS 2.1.2. The licensee stated that 
the proposed changes would be consistent with the NRC's Improved 
Standard Technical Specifications, NUREG-1433, and with NRC-approved 
(Safety Evaluation, dated May 15, 1991) Boiling Water Reactor Owners' 
Group Licensing Topical Report NEDO-31400A, dated July 9, 1987.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 1 in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the functions being removed do not contribute to avoidance 
of any previously evaluated accidents. Further, the changes have 
been shown to have an insignificant impact on overall reactivity 
control failure frequency. This insignificant impact is offset by 
the relatively large reduction in core damage frequency realized by 
the implementation of these changes. Hence, the probability or 
consequences of previously evaluated accidents are not significantly 
increased due to this change. To the contrary, as stated in the 
topical report [NEDO-31400A] the changes provide a net improvement 
in overall plant safety.
    The proposed amendment does not involve a physical or procedural 
change to any structure, component or system that significantly 
affects the probability or consequences of any accident or 
malfunction of equipment important to safety previously evaluated in 
the Final Safety Analysis Report (Updated). The proposed amendment 
will involve a change to reactor protection and isolation actuation 
systems circuitry that will remove the automatic reactor shutdown 
and Main Steam Line Isolation Valve closure functions of the Main 
Steam Line Radiation Monitor. However, the physical changes will not 
affect the remaining scram or vessel isolation functions.
    [***T]he methods, procedures and assumptions used to perform the 
eneric analyses in NED0-31400A are bounding for the Nine Mile Point 
Unit 1 with regard to input values. Niagara Mohawk has also provided 
in the evaluation reasonable assurance that significantly increased 
levels of radioactivity in the main steam lines will be controlled 
expeditiously to limit both occupational and environmental 
exposures. The Main Steam Line Radiation Monitor alarm setpoints 
will be set at 1.5 times the normal full power background dose rate 
and should any monitor exceed its alarm setpoint, the reactor 
coolant will be sampled to determine activity levels and the 
possible need for additional corrective actions.
    The offgas radiation monitor is a more sensitive monitor than 
the Main Steam Line Radiation Monitor because the nitrogen-16 
source, dominating the radiation signal to the Main Steam Line 
Radiation Monitor, has decayed by the time the radiation monitor can 
be affected by any increased levels of activity. Therefore, setting 
the offgas radiation monitor at 1.5 times the nitrogen-16 background 
dose rate is not reasonable since setting the monitor that low can 
lead to spurious activations of the alarm.
    Nine Mile Point Unit 1's monitor configuration, as described in 
the FSAR, detects the concentration of the offgas as it flows 
through the pipe. Thus, the detector is sensitive to fluctuations in 
condenser air inleakage, which can have an appreciable impact on the 
monitor readings, especially at readings as low as 1.5 times the 
normal full power background. Therefore, Niagara Mohawk proposes to 
set the alarm at five (5) times the normal full power background, 
which is still very conservative compared to the value allowed by 
Technical Specification 3.6.15.c., which is set based on Nine Mile 
Point Unit 1's Offsite Dose Calculation Manual.
    Niagara Mohawk believes that a setting of five (5) times the 
normal full power background is extremely conservative and is low 
enough to ensure detection of even minor fuel performance changes. 
Furthermore, if the monitor alarms at this setpoint of five times 
the normal full power background, the offgas will immediately be 
sampled and analyzed, followed by an analysis of a reactor coolant 
sample.
    Furthermore, the analyses in the Licensing Topical Report 
demonstrate that removal of the automatic reactor scram and Main 
Steam Line Isolation Valve closure functions of the Main Steam Line 
Radiation Monitor does not change the conclusions in the Final 
Safety Analysis Report (Updated) that the calculated radiological 
release consequences of the bounding control rod drop accident will 
not exceed the acceptable dose limits specified in 10CFR[Part]100.
    Therefore, Niagara Mohawk concludes that the proposed amendment 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 1 in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The function of a Main Steam Line Radiation Monitor trip is to 
detect abnormal fission produce release and isolate the main steam 
lines, thereby stopping the transport of fission products from the 
reactor to the main condenser. The monitors do not perform a 
prevention function for any kind of accident.
    The main steam line high radiation scram and main steam line 
isolation functions were originally intended to mitigate, not 
prevent, an existing accident scenario. However, the functions being 
removed do not contribute to avoidance or mitigation of any 
previously evaluated accidents since no credit is taken for these 
functions in any design basis event for terminating the initiating 
event or assuring the radioactive release remains within accepted 
limits. The existence of a Main Steam Line Radiation Monitor trip 
does not prevent the occurrence of a fuel failure event or any other 
type of event. Elimination of these functions will not introduce a 
new or different accident scenario.
    The proposed amendment represents a change to the physical 
configuration of the plant in that some reactor protection system 
circuits will be modified to eliminate the main steam line high 
radiation scram and main steam line isolation signals. However, 
these changes will not affect the remaining scram or vessel 
isolation functions. In all other respects, plant design and 
operation remain unchanged.
    Therefore, Niagara Mohawk Power Corporation concludes that the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The operation of Nine Mile Point Unit 1 in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety because, as shown in the topical report, the 
changes represent an overall improvement in plant safety in that the 
core damage frequency is reduced. Safe operation of the plant is 
enhanced by elimination of the unnecessary scram and isolation of 
the reactor vessel. With implementation of these changes, the 
primary heat sink remains available, a large transient on the vessel 
and safety-related actuations is avoided, and the Offgas System 
remains available to control the pathway of a potential release. 
Therefore, Niagara Mohawk concludes that the proposed amendment will 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: January 14, 1994
    Description of amendment request: The proposed amendment corrects 
an editorial error. Specifically, the amendment changes the reference 
in Limiting Condition for Operation (LCO) 3.4.D from ``3.3.A through 
C'' to ``3.4.A, 3.4.B, and 3.4.C.'' The amendment also changes the 
associated bases to clarify the LCO minimum solution concentration 
requirement of 11 weight percent and updates the excerpt from 10 CFR 
50.62 to reflect the current text of the regulation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    In accordance with 10 CFR 50.92, NNECO [Northeast Nuclear Energy 
Company] has reviewed the proposed change and has concluded that it 
does not involve a significant hazards consideration (SHC). The 
basis for this conclusion is that the three criteria of 10 CFR 
50.92(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 evaluated.
    The proposed change only modifies an incorrect reference in 
Section 3.4.D of the Technical Specifications. In practice, if 
Specification 3.4.A, 3.4.B, or 3.4.C cannot be met, an orderly 
shutdown is initiated. As currently written, the failure to meet the 
requirements of Section 3.3 would also initiate a shutdown in 
accordance with Section 3.4.D. This is not the intent of Section 
3.4.D since Section 3.3 already has specific shutdown requirements. 
This proposed change will correct Section 3.4.D so that it limits 
the conditions under which a plant shutdown must be initiated to the 
LCOs of the standby liquid control system. Therefore, this proposed 
change will not increase the probability or consequences of an 
accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed change corrects an incorrect section reference. 
There is no change to the operation or design of the plant, nor is 
there any change to the operability requirements of either section. 
The proposed change properly identifies the conditions under which 
the plant must be shutdown if an LCO is not met for the standby 
liquid control system. In practice, if Specification 3.4.A, 3.4.B, 
or 3.4.C cannot be met, an orderly shutdown is initiated. Since 
there is no change in plant operation or design, there is no 
possibility of a different kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not modify the design or function of 
the plant, nor does it reduce operability requirements of either 
Section 3.3 or 3.4. The proposed change only corrects an incorrect 
section reference by identifying the correct shutdown requirements 
for the standby liquid control system. Since there is no change to 
plant operation or design and the shutdown requirements are not 
reduced, there is 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: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz

Northeast Nuclear Energy Company (NNECO), Docket Nos. 50-245, 50-
336 and 50-423, Millstone Nuclear Power Station, Units 1, 2 and 3, 
New London County, Connecticut

    Date of amendment request: December 22, 1993
    Description of amendment request: The proposed amendments would 
change the Technical Specification (TS) as follows:
    1. Change the title of the Nuclear Station Director to Senior Vice 
President - Millstone Station.
    2. Remove the requirement to provide a copy of Plant Operations 
Review Committee (PORC) and Site Operations Review Committee (SORC) 
meeting minutes to the Executive Vice President - Nuclear. The Senior 
Vice President - Millstone Station is being proposed to replace the 
Executive Vice President - Nuclear for receipt of PORC and SORC meeting 
minutes.
    3. Make editorial changes to the Millstone Unit No. 1 TS Index.
    4. Correct a typographical error in Section 6.2.1.d of the 
Millstone Unit No. 1 TS.
    5. Correct a typographical error in Section 6.5.3.1.a of the 
Millstone Unit No. 3 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 (SHC), which is provided below:
    The proposed changes do not involve an SHC because the changes 
do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No design basis accidents are affected by these proposed 
changes. The proposed changes are administrative and editorial in 
nature to reflect a recent reorganization, removal of the Executive 
Vice President - Nuclear from receipt of PORC and SORC meeting 
minutes, addition of the Senior Vice President - Millstone Station 
to the receipt of PORC and SORC meeting minutes, and editorial 
changes to the Millstone Unit Nos. 1 and 3 Technical Specifications. 
No safety systems are adversely affected by the proposed changes, 
and no failure modes are associated with the changes. Therefore, 
there is no impact on the probability of occurrence or the 
consequences of any design basis events.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Since there are no changes in the way the plant is operated, the 
potential for an unanalyzed accident is not created. There is no 
impact on plant response, and no new failure modes are introduced. 
These proposed administrative and editorial changes have no impact 
on safety limits or design basis accidents, and they have no 
potential to create a new or unanalyzed event.
    3. Involve a significant reduction in a margin of safety.
    The changes do not directly affect any protective boundaries nor 
do they impact the safety limits for the protective boundaries. 
These proposed changes are administrative and editorial in nature. 
Therefore, there can be 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: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: December 17, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow a relaxation in setpoint 
tolerance of the pressurizer safety valves (PSVs) and main steam safety 
valves (MSSVs) from plus or minus 1% to plus or minus 3% for the ``as-
found'' test condition.
    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 (SHC), 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 evaluated.
    The proposed changes revise the ``as found'' setpoint tolerances 
for the PSVs and MSSVs from [plus or minus] 1% to [plus or minus] 
3%. For the resetting of the PSVs and MSSVs, a [plus or minus] 1% 
setpoint tolerance will be required prior to declaring the valve 
operable for those instances where the [plus or minus] 1% tolerance 
was exceeded. The proposed changes involve no hardware modifications 
to plant structures, systems, or components. The proposed setpoint 
tolerance of [plus or minus] 3% for the ``as-found'' condition was 
previously evaluated as part of the PSE [Plant Safety Evaluation] 
report for the transition to VANTAGE 5H fuel. The PSE was reviewed 
and approved by the NRC staff as a part of a prior license 
amendment.(9) In addition, since the proposed changes have 
previously been evaluated by the PSE report, the calculated 
radiological release associated with the PSE remain unaffected. In 
addition, the proposed changes are in compliance with applicable 
sections of the ASME Code and will not significantly affect 
structural integrity of either the reactor coolant system or the 
main steam system. Therefore, the proposed changes will have no 
effect on the probability or consequences of previously evaluated 
accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not create the possibility of a new or 
different kind of accident from those previously analyzed. The 
changes revise the Technical Specifications so that setpoint 
tolerance for the PSVs and MSSVs can be [plus or minus] 3% for the 
``as-found'' condition. These changes have no effect on plant 
operation. The PSV and MSSV setpoint drift in excess of the [plus or 
minus] 1% lift setting is an occurrence which has previously and may 
subsequently occur. The analyses for the transition to the VANTAGE 
5H fuel have examined the effects on the plant accident analyses for 
relaxation in PSV and MSSV setpoint tolerance to [plus or minus] 3%. 
Also, these changes will have no effect on ASME Code compliance. 
These changes do not introduce any new failures.
    3. Involve a significant reduction in the margin of safety.
    In support of the transition to the VANTAGE 5H fuel, a PSE was 
performed which assumed a [plus or minus] 3% setpoint tolerance for 
both the PSVs and MSSVs. Therefore, the effects of relaxing the PSV 
and MSSV setpoints are already accounted for in the existing 
analyses of record and will not affect the plants accident analyses. 
Additionally, the proposed changes will have no significant effect 
on the structural integrity of the reactor coolant system or the 
main steam system. Also, for those occurrences where the ``as-
found'' setpoint of the PSV or MSSV is in excess of [plus or minus] 
1%, a resetting to within [plus or minus] 1% of the valve setpoint 
will be required prior to declaring the valve operable. Therefore, 
the proposed changes will not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment requests: December 14, 1993 (Reference LAR 93-07)
     T3Description of amendment requests: The proposed amendment would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant (DCPP) Unit Nos. 1 and 2 to revise Technical Specification 
(TS) 3/4.8.1, ``A.C. Sources'' to increase the required quantity of 
emergency diesel generator (EDG) fuel oil stored in the engine-mounted 
tank (day tank). The amendment request also proposes to revise TS 3/
4.7.11, ``Area Temperature Monitoring,'' and 3/4.8.1 to remove 
references to a five EDG configuration. The specific TS changes 
proposed are as follows:
    (1) TS 3/4.7.11 would be revised to remove references to a common 
(swing) diesel generator in Table 3.7-5.
    (2) TS 3.8.1.1 and TS 3.8.1.2 would be revised to increase the 
required minimum contained volume in the EDG engine-mounted fuel tank 
(day tank) from 200 gallons to 250 gallons.
    (3) TS 3.8.1.1 and TS 4.8.1.1.2 would be revised to remove 
references to a five EDG configuration.
    (4) TS 3.8.1.2 would be revised to correct a footnote. TS Bases 3/
4.8.1, 3/4.8.2, and 3/4.8.3 would be revised to clarify commitments to 
Regulatory Guide 1.137 and expand the scope of information contained 
within the TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed increase in day tank TS minimum contained volume 
requirements provides additional conservatism to assure the EDG fuel 
oil contained in the day tank is sufficient to provide adequate time 
for an operator to take corrective action to restore the fuel oil 
supply to the affected day tank in the unlikely event that the fuel 
oil supply from the main tanks were cut off.
    Deletion of TS references to a five diesel generator 
configuration and correction of the TS 3.8.1.2 footnote are 
administrative changes that do not change the operating methodology 
of DCPP. These proposed administrative changes remove outdated 
information and correct an administrative oversight.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed increase in day tank TS minimum contained volume 
requirements would not involve any physical change to the plant 
systems or, in particular, to the EDG day tanks. The change does not 
affect the ability of the EDGs to start and to fulfill their safety-
related function. Hence, no new failure mechanisms will be 
introduced.
    The proposed removal of references to a five EDG configuration 
and correction of the TS 3.8.1.2 footnote are administrative in 
nature. Further, the proposed changes would not result in any 
physical alteration to any plant system. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Increasing the day tank TS minimum contained volume requirements 
is a conservative change which provides additional margin to assure 
the EDG fuel oil contained in the day tank is sufficient to provide 
adequate time for an operator to take corrective action to restore 
the fuel oil supply to the affected day tank in the unlikely event 
that the fuel oil supply from the main tanks were cut off. The 
proposed change will not alter any accident analysis assumptions, 
initial conditions, or results. Consequently, the proposed change to 
increase the EDG day tank TS contained fuel oil requirement does not 
have any effect on the margin of safety.
    The proposed administrative changes clarify the TS by removing 
references to a five diesel generator configuration and correcting 
the TS 3.8.1.2 footnote.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 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 NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: December 9, 1993
    Description of amendment request: The amendment would change the 
Operating Licenses and their corresponding Appendices A to reflect the 
planned implementation of the Power Rerate Program at Limerick 
Generating Station Units 1 and 2, and the corresponding increase in the 
authorized maximum reactor core power level by five percent to 3458 
megawatts thermal (MWt) from the current limit of 3293 MWt.
    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 Operation License (OL) changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed power rerate imposes only minor increases in the 
plant operating conditions. Plant systems, components, and 
structures have been verified to be capable of performing their 
intended functions under rerated conditions. Where necessary, some 
components will be modified or replaced prior to implementation of 
the Power Rerate Program to accommodate the revised operating 
condition. No new component or system interactions that could lead 
to an accident are created. As discussed below, no transient events 
result in a new sequence of events which could lead to a new 
accident scenario. Emergency Core Cooling Systems (ECCS) - Loss-of-
Coolant Accident (LOCA) Analysis.
    The current ECCS-LOCA performance analysis is already bounding 
for power rerate conditions. The fuel peak cladding temperature for 
rerate conditions is 1,345 deg.F, which is below the 2,200 deg.F 
regulatory limit. Therefore, the analysis demonstrates that the LGS, 
Units 1 and 2 will continue to comply with 10CFR50.46 and 10CFR50, 
Appendix K.
    Transient Event Analysis
    The evaluation results for transient events indicate the margin 
to the fuel Safety Limit Minimum Critical Power ratio (MCPR) will be 
maintained for the 8x8 array fuel types, such as GE8x8NB or GE11 
fuel design. The current fuel thermal-mechanical limits will 
continue to be met.
    Also, the power-dependent and flow-dependent MCPR and Maximum 
Average Planar Linear Heat Generation Rate (MAPLHGR) limits 
developed as part of the Average Power Range Monitor Rod Block 
Monitor Technical Specifications (ARTS) improvement program are 
applicable to power rerate. A TS Change Request to implement the 
ARTS improvement program was submitted to the NRC by letter dated 
August 27, 1993. The peak reactor vessel bottom head pressure will 
remain within the American Society for Mechanical Engineers (ASME) 
Code requirement for reactor overpressure protection.
    The analysis performed focused on the most limiting transient 
events in each disturbance category selected specifically for the 
power rerate evaluations. The results demonstrated that LGS, Unit 1 
and Unit 2 core thermal power output can be safely increased to 
power rerate parameters without impacting plant safety during a 
postulated transient event. The details of the impact to the 
description in the UFSAR are delineated below.
    a) Events Resulting in a Core Coolant Temperature Decrease
    i) Loss of Feedwater Heating (LFWH)
    The delta Critical Power Ratio (delta CPR) for the LFWH event at 
the rerated power is bounded by the result estimated for the current 
rated power level and remains significantly less than the Operating 
Limit MCPR. There is no change between the delta CPR results for 
high and low reactor core flow conditions. The calculated thermal 
and mechanical overpowers for this event at power rerate conditions 
also meet the fuel design criteria.
    ii) Feedwater Controller Failure (FWCF) Maximum Demand
    For the Increased Core Flow (ICF) and the Maximum Extended Load 
Line Limit (MELLL) conditions, the trend for the FWCF - Maximum 
Demand event at rerate conditions is consistent with the current 
rated power analysis. For both high and low reactor core flow 
conditions, the FWCF - Maximum Demand event becomes most limiting 
due to the Turbine Bypass Valve Out-of-Service (TBVOOS) and the 
Recirculation Pump Trip Out-of-Service (RPTOOS) analyses assumption. 
The fuel thermal margin results remain within the acceptable limits 
for the fuel type analyzed.
    b) Events Resulting in a Reactor Pressure Increase
    i) Turbine Trip with No Bypass (TTNBP)
    At rerate conditions, the fuel transient thermal and mechanical 
overpower results remain below the NRC acceptance criteria.
    ii) Generator Load Rejection with No Bypass (LRNBP)
    The fuel transient thermal responses are less severe than for 
the TTNBP event described above. Therefore, at power rerate 
conditions, the LRNBP event remains bounded by the TTNBP event.
    iii) Main Steam Isolation Valve Closure, Flux Scram (MSIVF)
    The peak reactor vessel bottom head pressure for rerate 
conditions is slightly higher than the pressure at current rated 
conditions due to the higher initial reactor coolant system 
pressure. However, this result is still below the ASME overpressure 
limit of 1,375 psig by a margin of 33 psi.
    c) Events Resulting in a Core Coolant System Flow Rate Decrease
    i) Recirculation Pump Seizure
    The recirculation pump seizure assumes instantaneous stoppage of 
the pump motor shaft of one recirculation pump. As a result, the 
reactor core flow decreases rapidly. The reactor flow decreases 
rapidly. The reactor vessel level swell due to the rapid reactor 
core flow reduction reaches the high reactor water level setpoint, 
causing a feedwater pump trip, a main turbine trip, and subsequently 
a reactor scram on turbine stop valves closure. The peak neutron 
flux and average fuel surface heat flux do not increase 
significantly above the initial conditions, therefore no impact on 
the fuel thermal margin is postulated to occur.
    d) Events Resulting in Reactivity and Power Distribution 
Anomalies
    i) Rod Withdrawal Error (RWE)
    The calculated delta CPR of 0.10 for this event at rerate 
conditions is bounded by the generic ARTS - based RWE limits of 
0.13. Therefore, the generic ARTS-based RWE analysis delta CPR 
result is verified to be applicable for power rerate conditions for 
LGS Units 1 and 2.
    e) Events Resulting in a Reactor Coolant Inventory Increase
    i) Inadvertent High Pressure Coolant Injection (HPCI) System 
Actuation
    Based on the peak average fuel surface heat flux results, the 
HPCI actuation event will be bounded by the limiting pressurization 
event (i.e., the TTNBP event described above) for delta CPR 
consideration.
    Anticipated Transients Without SCRAM (ATWS) Analysis
    A generic evaluation for the ATWS event is provided in Section 
3.7 of the Topical Report NEDC-31984P, ``Generic Evaluations of 
General Electric Boiling Water Reactor Power Uprate,'' Supplement 1, 
dated July 1991. This evaluation concludes that the ATWS acceptance 
criteria for fuel, reactor pressure vessel (RPV) and containment 
integrity will be met, if the following exists;
    - Reactor power increases less than or equal to 5%
    - Reactor Steam Dome pressure increases less than or equal to 40 
psi;
    - Safety Relief Valve (SRV) opening setpoints increase less than 
or equal to 80 psi; and
    - ATWS high pressure setpoint increases less than or equal to 20 
psi.
    The plant's parameter changes will remain within the above 
criteria, except that the ATWS high pressure setpoint increase is 40 
psi rather than 20 psi in order to maintain the same relationship 
between the ATWS high pressure setpoint and the SRV opening 
setpoints. Based on the previous analysis, this difference would 
have a minor effect on the analysis results. The only significant 
change is a slightly higher (i.e., about 10 psi) peak RPV pressure.
    For additional assurance, a LGS specific ATWS analysis for a 5% 
power rerate was performed. The events analyzed were:
    1. Main Steam Isolation Valve (MSIV) Closure,
    2. Pressure Regulator Failure - Open,
    3. Loss of Feedwater, and
    4. Inadvertent Opening of a Relief Valve.
    The LGS specific analysis also concludes that the ATWS 
acceptance criteria for fuel, RPV, and containment integrity will be 
met for a 5% power rerate.
    Other Evaluations
    The impact of power rerate on the radiological consequences of 
the accidents presented in UFSAR Chapter 15 was determined based on 
the current design basis analyses, post rerate implementation system 
conditions, and radiological source terms. In general, power rerate 
will result in a small increase in the quantity of radioactive 
material released during accidents and therefore slightly higher 
(i.e., approximately 2% to 5%) accident doses. However, USFAR 
Chapter 15 accident doses for rerated conditions remain within the 
regulatory limits specified in 10CFR100 and 10CFR50, Appendix A, GDC 
19.
    The UFSAR Chapter 15 accidents that were evaluated and updated 
for rerate conditions are as follows:
    1) Loss of Coolant Accident (LOCA)
    2) Main Steam Line Break (MSLB)
    3) Fuel Handling Accident
    4) Control Rod Drop Accident
    5) Instrument Line Break
    6) Feedwater Line Break
    7) Steam Jet Air Ejector Line Break
    8) Offgas System Failure
    9) Liquid Radioactive Waste System Failure
    An evaluation was also performed to address the power rerate 
impact on accident mitigative features, structures, systems, and 
components, within the balance of plant. The results are as follows:
    - Auxiliary systems such as the Emergency Service Water, 
Residual Heat Removal (RHR) Service Water, Ultimate Heat Sink (i.e., 
the spray pond), safety-related portions of secondary containment 
reactor enclosure air cooling, primary containment drywell air 
recirculation, and Emergency Diesel Generator enclosure ventilation 
were confirmed to operate acceptably under normal and accident 
conditions after implementation of power rerate.
    - Combustible gas control systems were confirmed to be capable 
of maintaining oxygen concentrations inside the primary containment 
within regulatory limits under post accident rerate conditions.
    - The secondary containment reactor enclosure recirculation 
system and Standby Gas Treatment system were confirmed to be able to 
adequately contain, process, and control the release of normal and 
post-accident levels of radioactive material after implementation of 
power rerate.
    - Instrumentation was reviewed and confirmed to be capable of 
performing their control and monitoring functions under rerate 
conditions.
    - Electric power systems including the main turbine generator 
and switchgear components were verified as being capable of 
providing the electrical load as a result of the rerated power 
levels. No safety-related electrical loads were affected which would 
impact the Emergency Diesel Generators.
    - Piping systems were evaluated for the effect of operation at 
higher power levels, including transient loadings. The evaluation 
confirmed that with few exceptions piping and supports are adequate 
to accommodate the increased loadings resulting from operation at 
rerated power conditions. In a few cases, piping supports will be 
modified to accept the higher forces due to rerate conditions.
    - The effect of rerate conditions on high energy line break 
(HELB) events for all Nuclear Steam Supply System (NSSS) and Balance 
of Plant (BOP) systems was evaluated. The evaluation confirmed 
structures, systems, and components important to safety are capable 
of accommodating the effects of jet impingement and blowdown forces 
and the environmental effects resulting from HELB events at rerate 
conditions.
    - The Moderate Energy Line Break (MELB) analysis was evaluated 
for impact due to rerate conditions. Sufficient margin was 
determined to exist in the original analysis to bound the rerate 
conditions.
    - Main control room (MCR) habitability was evaluated. Post-
accident MCR and Technical Support Center (TSC) doses were confirmed 
to be within the limits of General Design Criterion (GDC) 19 of 
10CFR50 Appendix A.
    - Radiation doses for normal operation were reviewed and 
confirmed to remain within the limits of 10CFR20 and 10CFR50, 
Appendix I. The impact on post-accident sampling activities and 
post-accident access to vital areas was also confirmed to be 
acceptable.
    - The environmental qualification of electrical and mechanical 
equipment important to safety was evaluated for the impact of normal 
and accident operating conditions at rerated power levels. The 
majority of equipment will remain qualified for the new conditions. 
For equipment that is not qualified, corrective actions will be 
taken to ensure the plant equipment will perform their intended 
functions under rerate conditions. No new equipment will be added 
for power rerate which would increase the potential for component 
failure. The Preventative Maintenance Program (PMP) will continue to 
provide for appropriate equipment repair or replacement during 
operation at rerated power conditions.
    - The impact of operation at rerated power levels was evaluated 
for Station Blackout and Fire Safety Shutdown area heat-up concerns. 
The evaluation confirmed there is no adverse impact from rerate on 
the ability of the plant to achieve safe shutdown under these 
conditions.
    - The consequences of postulated transients and special events 
(i.e., ATWS and Station Blackout) will remain within NRC acceptance 
criteria for rerate conditions. Concurrent malfunctions assumed to 
occur during accidents have been accounted for in the safety 
analyses for rerate conditions. The consequences of these equipment 
malfunctions will not change with implementation of the Power Rerate 
Program. Equipment that is important to safety either is capable of 
or will be modified and/or replaced to be capable of performing its 
intended function. The availability of redundant systems to provide 
safety functions in the event of component malfunction is not 
impacted as a result of rerate conditions. Furthermore, the impact 
of power rerate on the consequences of abnormal transients and 
accident conditions which are a result of component malfunctions has 
been shown to be acceptable.
    The probability (i.e., frequency of occurrence) of Design Basis 
Accidents (DBAs) occurring is not affected by the proposed increased 
power level, as the applicable regulatory criteria established for 
plant equipment (e.g., ASME Code, the Institute of Electrical and 
Electronics Engineers (IEEE) standards, National Electrical 
Manufacturer's Association (NEMA) standards, NRC Regulatory Guides) 
will still be followed as the plant is operated at the rerated power 
level. Reactor SCRAM setpoints will be established such that there 
is no significant increase in frequency due to rerate conditions. No 
new challenges to safety-related equipment will result from the 
implementation of power rerate.
    The changes in consequences of hypothetical accidents which 
would occur from 102% of the rerated power, compared to those 
previously evaluated, are in all cases not significant, because the 
accident evaluations from a power rerate to 105% of original rated 
power will not result in exceeding the applicable NRC approved 
acceptance limits. The spectrum of hypothetical accidents and 
transients has been investigated, and has been determined to meet 
the current regulatory criteria for LGS, Units 1 and 2 at rerate 
conditions. The offsite radiological doses resulting from DBAs are 
calculated to increase by only a few percent (i.e., approximately 2% 
to 5%) because of the rerated power level, and will remain below 
10CFR100 limits. In the area of reactor core design, the fuel 
operating limits will continue to be met at the rerated power level, 
and fuel reload analyses will continue to show that plant transients 
will meet the criteria accepted by the NRC as specified in NEDO-
24011, ``GESTAR II.''
    Challenges to fuel or ECCS performance were evaluated and shown 
to still meet the criteria of 10CFR50.46 and 10CFR50, Appendix K. 
Challenges to the primary containment have been evaluated and still 
meet 10CFR50, Appendix A, GDC 38, ``Long Term Cooling,'' and GDC 50, 
``Containment.'' Radiological release events have been evaluated and 
have been shown to meet the guidelines of 10CFR100.
    Therefore, the proposed OL changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2) The proposed OL changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    All actions to ensure that safety-related structures, systems, 
and components will remain within their design allowable values, and 
ensure that they can perform their intended functions under rerate 
conditions will be taken prior to implementation of power rerate. 
Power rerate does not increase challenges to or create any new 
challenges to safety-related equipment or other equipment whose 
failure could cause an accident. No new equipment is added as a 
result of implementing the Power Rerate Program which would create 
the possibility of a new type of accident. In addition, power rerate 
does not create any new sequence of events or failure modes that 
lead to a new type of accident.
    Implementation of power rerate will increase the average neutron 
flux in the reactor core, which increases the integrated neutron 
fluence on the reactor pressure vessel (RPV) wall. To account for 
the higher fluence, an RPV fracture toughness analysis was performed 
for power rerate conditions. This analysis resulted in a proposed 
revision to the ``pressure vs. temperature'' curves currently 
provided in the Technical Specifications (TS), that will maintain 
the current level of protection for the RPV. Therefore, power rerate 
will not result in any new failure mode for the RPV, and thus, does 
not create the possibility of a different type of accident from any 
accident previously evaluated.
    No new operating mode, safety-related equipment lineup, accident 
scenario, or equipment failure mode was identified as resulting from 
the implementation of the Power Rerate Program. The full spectrum of 
accident considerations defined in NRC Regulatory Guide 1.70, 
``Standard Format and Content of Safety Analysis Reports for Nuclear 
Power Plants - LWR Edition,'' Revision 3, dated November 1978, have 
been evaluated for rerate conditions and no new or different kind of 
accident has been identified. Implementation of the Power Rerate 
Program uses already-developed technology and applies it within the 
capabilities of already existing plant equipment in accordance with 
presently existing regulatory criteria to include applicable NRC 
approved codes, standards, and methods. General Electric (GE) has 
designed Boiling Water Reactors (BWRs) of higher power levels than 
the rerated power of any of the currently operating BWR fleet and no 
new power dependent accidents have been identified.
    Therefore, the proposed OL changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3) The proposed OL changes do not involve a significant 
reduction in a margin of safety.
    Power rerate will not involve a significant reduction in a 
margin of safety, as plant equipment and reactions to transients and 
hypothetical accidents will not result in exceeding the presently 
approved NRC acceptance limits. The accident doses are calculated to 
increase a few percent (approximately 2% to 5%) because of power 
rerate, but remain below 10 CFR 100 limits. The events (i.e., 
transients, accidents, and ATWS) that form the bases of the TS were 
evaluated for power rerate conditions. Although some changes to the 
TS are required to implement power rerate, no NRC acceptance limit 
will be exceeded. Therefore, the margins of safety with respect to 
the safety limits and other TS bases will be maintained.
    For systems addressed in the TS Section 2.2, 3/4.1, 3/4.2, 3/
4.3, 3/4.4, 3/4.5, 3/4.6 and 3/4.7 (i.e., Reactor Protection System, 
Standby Liquid Control System, Power Distribution Limits, 
Instrumentation, Reactor Coolant System, Emergency Core Cooling 
Systems, Containment Systems, and Plant Systems), all components 
will be operable and capable of performing their intended functions 
under power rerate conditions such that the margin of safety is not 
adversely impacted.
    Therefore, the proposed OL changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

Philadelphia Electric 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: December 21, 1993
    Description of amendment request: The proposed Technical 
Specification (TS) changes revise Table 3.2.F, ``Surveillance 
Instrumentation,'' to accurately describe the main stack high range and 
reactor building roof vent high range radiation monitors, and deletes 
previously approved TS Change Request (TSCR) 91-10 for Unit 3 (License 
Amendment No. 168). TSCR 91-10 requested an emergency temporary change 
to the TS to allow fuel loading to take place without all control rods 
fully inserted into the core.
    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.
    Because the proposed changes are administrative in nature, they 
do not affect the initial conditions or precursors assumed in the 
Updated Final Safety Analysis Report Section 14. These changes do 
not decrease the effectiveness of equipment relied upon to mitigate 
the previously evaluated accidents.
    Therefore, there is no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not make any physical changes to the 
plant or changes to operating procedures. Therefore, 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.
    Therefore, the proposed changes do 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 changes are administrative in nature and are 
intended to provide clarification or eliminate confusion when 
interpreting the Technical Specifications. The proposed changes do 
not adversely affect the assumptions or sequence of events used in 
any accident analysis.
    Therefore, the proposed changes do not involve a reduction in 
any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105

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

    Date of amendment request: October 29, 1993
    Description of amendment request: The licensee is preparing to 
replace the currently installed steam generators with new model Delta 
75 steam generators (Delta 75 SGs). The new steam generators will be 
larger than those currently installed. The physical changes to the 
plant and the accident reanalyses needed to support those changes will 
necessitate changes to the Technical Specifications (TS). The TS 
changes requested involve alterations to the core operating limits, 
changes to various reactor trip setpoints, deletion of the negative 
flux rate trip, removal of references to specific analyses, changes to 
the steam/feedwater flow mismatch activation setpoint, changes to 
shutdown limits, changes to instrument uncertainty allowances, a change 
to the methodology for reactor coolant system (RCS) flow determination, 
modifications to departure from nucleate boiling (DNB) parameters, a 
change to the engineered safety features actuation system setpoints for 
steam generator water levels, removal of the F* and L* criteria, and 
the addition of a requirement for a first inservice inspection for the 
new steam generators. Due to the size of the new steam generators, TS 
containing references to the maximum containment pressure following a 
steam line break and the total RCS volume will also change; in 
addition, a reference to RCS temperature is changed from a nominal 
value to an indicated value.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), South Carolina Electric 
& Gas Company (SCE&G or the licensee) has provided its analysis of the 
issue of no significant hazards consideration, which is presented 
below:
    1) Operation of VCSNS [Virgil C. Summer Nuclear Station] in 
accordance with the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Implementation of the [Delta] 75 SGs and revised operating 
conditions do not contribute to the initiation of any accident 
evaluated in the FSAR [Final Safety Analysis Report]. Supporting 
factors are as follows:
    - The [Delta] 75 SG is designed in accordance with ASME 
[American Society of Mechanical Engineers] Code Section III, 1986 
edition [sic] and other applicable federal, state, and local laws, 
codes and regulations and meets the original interfaces for the 
Model D3 SGs with exception that provisions for a larger blowdown 
nozzle have been made and the feedwater inlet nozzle is located in 
the upper shell.
    - All NSSS [nuclear steam supply system] components (i.e., 
reactor vessel, RC Pumps, pressurizer, CRDM's [control rod drive 
mechanisms], [Delta] 75 SGs, and RCS piping) are compatible with the 
revised operating conditions. Their structural integrity is 
maintained during all proposed plant conditions through compliance 
with the ASME code.
    - Fluid and auxiliary systems which are important to safety are 
not adversely impacted and will continue to perform their design 
function.
    - Overall plant performance and operation are not significantly 
altered by the proposed changes.
    Therefore, since the reactor coolant pressure boundary integrity 
and system functions are not adversely impacted, the probability of 
occurrence of an accident evaluated in the VCSNS FSAR will be no 
greater than the original design basis of the plant.
    An extensive analysis has been performed to evaluate the 
consequences of the following accident types currently evaluated in 
the VCSNS FSAR:
    - Non-LOCA [loss-of-coolant accident]
    - Large Break LOCA
    - Steam Generator Tube Rupture
    With the [Delta] 75 SGs and revised operating conditions, the 
calculated results (i.e., DNBR [departure from nucleate boiling 
ratio], Primary and Secondary System Pressure, Peak Clad 
Temperature, Metal Water Reaction, Challenge to Long Term Cooling, 
Environmental Conditions Inside and Outside Containment, etc.) for 
the accidents are similar to those currently reported in the VCSNS 
FSAR. Select results (i.e., Containment Pressure During a Steam Line 
Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are 
slightly more limiting than those reported in the current FSAR due 
to the use of the assumed operating conditions with the new [Delta] 
75 SGs, and in some cases, use of an uprated core power of 2900 MWt. 
However, in all cases, the calculated results do not challenge the 
integrity of the primary/secondary/ containment pressure boundary 
and remain within the regulatory acceptance criteria applied to 
VCSNS's current licensing basis. The assumptions utilized in the 
radiological evaluations, described in Section 3.7, are thus 
appropriate and are judged to provide a conservative estimate of the 
radiological consequences during accident conditions. Given that 
calculated radiological consequences are not significantly higher 
than current FSAR results and remain well within 10CFR100 limits, it 
is concluded that the consequences of an accident previously 
evaluated in the FSAR are not increased.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The [Delta] 75 SGs and revised operating conditions will not 
introduce any new accident initiator mechanisms. Structural 
integrity of the RCS is maintained during all plant conditions 
through compliance with the ASME code. No new failure modes or 
limiting single failures have been identified. Design requirements 
of auxiliary systems are met with the RSGs [Replacement Steam 
Generators]. Since the safety and design requirements continue to be 
met and the integrity of the reactor coolant system pressure 
boundary is not challenged, no new accident scenarios have been 
created. Therefore, the types of accidents defined in the FSAR 
continue to represent the credible spectrum of events to be analyzed 
which determine safe plant operation.
    3) The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    Although the [Delta] 75 SGs and revised operating conditions 
will require changes to the VCSNS Technical Specifications, it will 
not invalidate the LOCA, non-LOCA, or SGTR [steam generator tube 
rupture] conclusions presented in the FSAR accident analyses 
(Appendix 6). For all the FSAR non-LOCA transients, the DNB design 
basis, primary and secondary pressure limits, and dose limits 
continue to be met. The LOCA peak cladding temperatures remain below 
the limits specified in 10CFR50.46. The calculated doses resulting 
from a SGTR event will continue to remain within a small fraction of 
the 10CFR100 permissible releases. Environmental conditions 
associated with High Energy Line Break (HELB) both inside and 
outside containment have been evaluated. The containment design 
pressure will not be violated as a result of the HELB. Equipment 
qualification will be updated, as necessary, to reflect the revised 
conditions resulting from HELB. The margin of safety with respect to 
primary pressure boundary is provided, in part, by the safety 
factors included in the ASME Code. Since the components remain in 
compliance with the codes and standards in effect when VCSNS was 
originally licensed (with the exception of the [Delta] 75 RSGs which 
use the 1986 ASME Code Section III Edition), the margin of safety is 
not reduced. Thus, there is no reduction in the margin to safety as 
defined in the bases of the VCSNS 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: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: S. Singh Bajwa

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

    Date of amendment request: December 17, 1993
    Description of amendment request: The proposed changes would revise 
Technical Specification 3/4.3.3.6, ``Accident Monitoring 
Instrumentation,'' and the associated Technical Specification Bases. 
The changes are in accordance with the applicable guidance of Revision 
3 to Regulatory Guide (RG) 1.97.
    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 would 
not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Regulatory Guide 1.97 furnishes standards acceptable to the NRC 
for instrumentation to monitor plant variables and systems during 
and following an accident. The purpose of the accident monitoring 
instrumentation is to display plant variables that provide 
information required by the control room operators for manual 
actions and long term recovery. Determination of variable types and 
category designations for VCSNS [Virgil C. Summer Nuclear Station] 
was accomplished from a review of the Emergency Response Guidelines 
(ERGs), the Final Safety Analysis Report, and the Westinghouse 
Owners Group (WOG) ERGs. The WOG ERGs were used at VCSNS as a basis 
for the Emergency Response Procedures. Operability of the 
instruments used for accident monitoring ensures there is sufficient 
information available on selected plant parameters to monitor plant 
status during and following an accident. The changes proposed do not 
effect components that can cause an accident. The increase in 
allowable outage times from 7 to 30 days or from 48 hours to 7 days 
does not significantly affect the consequences of an event 
previously evaluated. The channel redundancy and the relatively 
short outage times, coupled with the low probability of an event 
requiring accident monitoring instrumentation during this interval, 
ensure that sufficient information is available for operator manual 
actions. The condition of the plant in either HOT STANDBY or HOT 
SHUTDOWN, the first stage of the plant shutdown process, has no 
impact on the assumptions made in the accident analysis.
    The change in mode applicability for the Reactor Building Area 
High Range Radiation Monitors to include modes 1, 2, and 3, but 
exclude mode 4, is based on the usage of these monitors which is to 
indicate a significant degradation of the reactor coolant pressure 
boundary. These monitors do not initiate any automatic mitigation 
system and are solely required to be operable to provide indication 
which in conjunction with other operator actions will aid in 
mitigating the consequences of design basis accidents. Design basis 
accident sequences which may create a significant degradation of the 
reactor coolant pressure boundary are not postulated to occur during 
mode 4. Therefore, the proposed change does not increase the 
probability or consequences of any accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change is consistent with the requirements of RG 
1.97. The accident monitoring instrumentation will make available 
reliable information to plant control room operators to mitigate the 
consequences of a design basis accident. The first stages of plant 
shutdown, HOT STANDBY and HOT SHUTDOWN, are plant modes for which 
VCSNS has been analyzed. Since no plant configuration changes or 
changes to the mode of operation of equipment, systems, and 
components are introduced by the proposed Technical Specification, 
no new failure modes or accident sequences are instituted. 
Therefore, the changes proposed do not create the possibility of a 
new or different kind of accident from any previously analyzed.
    (3) Involve a significant reduction in a margin of safety.
    The inclusion of category 1, type A or B, instrumentation in the 
TS [Technical Specifications] provides assurance that adequate 
information is available to the operators to maintain VCSNS in a 
safe condition during and following a design basis accident. 
Accomplishment of specific manual action by the control room 
operators is enhanced due to the availability and reliability of the 
indications. The proposed changes do not affect the design or 
operation of safety related components relied upon to automatically 
mitigate the consequences of a design basis event. The proposed 
change from HOT SHUTDOWN to HOT STANDBY as the first stage of plant 
shutdown will not affect the design or operation of any safety 
related system or component. Therefore, the changes proposed would 
not involve a reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: S. Singh Bajwa

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

    Date of amendment requests: November 3, 1993
     T3Description of amendment requests: The licensee proposes to 
revise the operability requirements of containment isolation valves 
listed in Technical Specification (TS) Table 3.6-1, Section D. The 
associated Bases 3/5.6.3, ``Containment Isolation Valves,'' is also 
revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    The proposed change provides new actions and Allowed Outage 
Times (AOTs) for valves in Section D of Technical Specification (TS) 
Table 3.6-1 that are currently allowed by the existing TS to be 
secured for an indefinite period of time as long as they are secured 
in their Engineered Safety Feature Actuation System (ESFAS) actuated 
position. These valves are considered operable by the existing TS 
although they may be unable to perform their containment isolation 
function. The proposed change ensures that these valves are returned 
to operable status within specified times based on the results of 
specific risk evaluations on their contribution to core damage or 
offsite dose release. The proposed change does not involve a 
physical change to the facility as described in the Updated Final 
Safety Analysis Report (UFSAR). Therefore, this proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No
    The ESFAS actuated positions of these valves are the positions 
assumed in the safety analysis. There are no new accidents 
associated with this proposed change because the previously analyzed 
events already considered failures of containment isolation valves. 
The plant is equipped with dual and redundant containment isolation 
valves. Leaving the valves in their ESFAS actuated positions does 
not create a new accident. Therefore, this proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No
    This proposed change 1) limits the AOT of certain valves based 
on contributions to core damage and offsite dose release when the 
valves are secured in their ESFAS actuated position and 2) requires 
these valves to be returned to OPERABLE status prior to Mode 4 entry 
from a cold shutdown to ensure they are available to perform their 
intended containment isolation function. Previously, these valves 
could be secured in the ESFAS actuated position indefinitely. 
Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713
    Attorney for licensee: James A. Beoletto, Esquire, Southern 
California Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: Theodore R. Quay

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: December 27, 1993
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Units No. 1 and No. 2 (NA-1&2). The proposed changes revise the review 
responsibilities of the Station Nuclear Safety and Operating Committee 
(SNSOC) and the Management Safety Review Committee (MSRC).
    The NA-1&2 TS address the organization and responsibilities of both 
the onsite and offsite review groups: SNSOC and MSRC, respectively. The 
responsibilities of the SNSOC include the review of new procedures and 
changes to procedures that affect nuclear safety. The MSRC review 
responsibilities include the review of safety evaluations and SNSOC 
meeting minutes and reports. The extent of these review activities 
would be revised by the proposed changes to ensure the two review 
groups are focusing on nuclear safety issues and not spending an 
unnecessary amount of time on activities of minimal safety 
significance. Specifically, the proposed changes would revise the 
review responsibilities of SNSOC regarding procedure changes. Rather 
than reviewing all procedure changes, SNSOC would only review procedure 
changes that require a safety evaluation. The proposed changes also 
would revise the review responsibilities of the MSRC. Rather than 
reviewing all of the safety evaluations and SNSOC meeting minutes and 
reports as presently required by the TS, the MSRC would only review a 
representative sample of these documents.
    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:
    [T]he elimination of the SNSOC review of procedure changes that 
do not require a safety evaluation, revising the wording for 
approval of procedure changes, and the modification of the MSRC's 
duties regarding their review of safety evaluations and SNSOC 
meeting minutes and reports will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. As administrative 
changes, the proposed Technical Specifications changes have no 
direct or indirect effect on accident precursors. No plant 
modifications are being implemented and operation of the plant is 
unchanged. SNSOC review of new procedures and procedure changes that 
require a safety evaluation ensures that activities that could 
affect nuclear safety are being properly reviewed. The MSRC's 
overview of representative samples of safety evaluations and SNSOC 
meeting minutes and reports based on performance ensures these 
programs are being properly implemented and nuclear safety is not 
being compromised; or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated since physical modifications 
are not involved and systems and components will be operated as 
before the change. The proposed changes are wholly administrative in 
nature and have no impact on plant operations or accident 
considerations. These changes modify the scope of SNSOC review of 
procedure changes and MSRC's review functions concerning safety 
evaluations and SNSOC meeting minutes and reports. Procedure changes 
will continue to receive management review in accordance with 
administative procedures, however, only changes that require a 
safety evaluation will require SNSOC approval. MSRC review of 
representative samples of safety evaluations and SNSOC meeting 
minutes and reports based on performance will continue to provide 
adequate assurance that nuclear safety is being properly considered; 
or
    3. Involve a significant reduction in a margin of safety as 
defined in the basis of any Technical Specification since the 
responsibilities of the SNSOC and MSRC are not addressed by the 
existing Technical Specification Bases, nor are review requirements 
for procedures. The proposed changes are administrative in nature 
and have no impact on, nor were they considered in, existing UFSAR 
accident analyses. Safety significant procedure changes, i.e., 
changes that require a safety evaluation to be prepared, will 
continue to be reviewed by SNSOC, as will new procedures. Procedure 
changes still require cognizant management approval and preparation 
of an activity screening to determine whether or not the change 
impacts nuclear safety. This ensures activities important to nuclear 
safety are being appropriately reviewed. The effectiveness of the 
safety evaluation program, and the thoroughness of SNSOC meetings 
and reports will be assured through the MSRC's plant overview 
function which is based on observed performance.
    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: Herbert N. Berkow

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.

Duke Power Company, Docket No. 50-413, Catawba Nuclear Station, 
Unit No. 1, York County, South Carolina

    Date of amendment request: January 10, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 2.0 and 3/4.2 which currently requires 
the determination of the reactor coolant system flow rate by precision 
heat balance measurement at least once per 18 months. Date of 
publication of individual notice in Federal Register: January 26, 1994 
(59 FR 3743)
    Expiration date of individual notice: February 25, 1994
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of amendment request: November 11, 1993, as supplemented 
November 22, 1993
    Description of amendment request: The proposed amendments would 
provide an interim acceptance criteria for control rod drop time on 
Oconee Unit 1. Specifically, control rod Group 1, Rod 8, and Group 2, 
Rod 5, would be considered operable with an insertion time of less than 
or equal to 3.00 seconds provided that: (1) the average insertion time 
for the remaining rods in Group 1 and the average insertion time for 
the remaining rods in Group 2 is less than or equal to 1.5 seconds, and 
(2) the core average negative reactivity insertion rate is within the 
assumptions of the safety analysis. The acceptance criteria would apply 
until the end of the current fuel cycle for Oconee Unit 1. This 
acceptance criteria for rod drop time would apply for the two rods, 
rather than the existing Technical Specification 4.7.1 limit of 2.00 
seconds from the fully withdrawn position to 3/4 insertion.Date of 
publication of individual notice in Federal Register: November 29, 1993 
(58 FR 62689)
    Expiration date of individual notice: December 29, 1993
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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

    Date of application for amendment: December 8, 
1993I11T3Brief description of amendment request: The 
proposed amendment would grant one-time extensions for certain 
Technical Specification surveillances which are currently required to 
be performed beginning February 16, 1994. The licensee is requesting 
extension of the surveillance intervals because the current operating 
cycle has been extended, impacting the required completion dates for 
these surveillances. Performance of these surveillances within the 
required intervals would require that the plant be placed in an 
undesirable operating configuration, or would necessitate a plant 
shutdown. The surveillances for which extensions have been requested 
will be performed during the fifth refueling outage, scheduled to begin 
on April 16, 1994.
    Date of individual notice in Federal Register: January 18, 1994 (59 
FR 2630)
    Expiration date of individual notice: February 17, 1994
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

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: December 22, 1993
    Description of amendment request: The proposed amendment would add 
Limiting Conditions for Operation (LCO) and Surveillance Requirements 
to Tables 3.12.1, ``Water Spray/Sprinkler Protected Areas,'' and 
4.12.1, ``Water Spray/Sprinkler Tests,'' and clarify the associated 
Bases to reflect the installation of a new full area fire suppression 
system in the east and west cable tunnels. This new full area fire 
suppression system was installed because the previous sprinkler system 
did not provide coverage to some cable trays and the sprinkler head 
orientation did not provide full coverage of the cable trays where it 
was installed. The proposed amendment would also correct other portions 
of Tables 3.12.1 and 4.12.1 for consistency with changes made to 
reflect the east and west cable tunnel modification.Date of publication 
of individual notice in Federal Register: January 18, 1994 (59 FR 2634)
    Expiration date of individual notice: February 17, 1994
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.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 20555, and at the local public document 
rooms for the particular facilities involved.

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: October 28, 1992, as 
supplemented December 14, 1993
    Brief description of amendments: The amendments remove Table 4.4-5, 
``Reactor Vessel Material Surveillance Program Withdrawal Schedule,'' 
from the McGuire Technical Specifications and make other administrative 
changes associated with the removal of the withdrawal schedule in 
accordance with NRC Generic Letter 91-01.
    Date of issuance: January 31, 1994
    Effective date: January 31, 1994
    Amendment Nos.: 139 and 121
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 23, 1992 (57 
FR 61112) The December 14, 1993, letter provided clarifying information 
that did not change the scope of the October 28, 1992, 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 January 31, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Astkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South 
CarolinaDate of application of amendments: July 14, 1993, as 
supplemented August 24 and September 22, 1993

    Brief description of amendments: The amendments revise TS 3.1.2.9 
to clarify the role of High Pressure Injection and Core Flood Tank 
deactivation in maintaining pilot operated relief valve operability for 
low temperature overpressure protection (LTOP), add restrictions 
regarding applicability of controls which assure 10 minutes are 
available for operator action to mitigate an LTOP event, revise the 
pressure-temperature limits and associated LTOP setpoints, and make 
associated administrative changes. Also, the Bases would be revised to 
be consistent with the above changes.
    The conformance of the upper shelf energy and reactor vessel 
material surveillance program to Appendices G and H will be determined 
pending the NRC staff resolution of Generic Letter 92-01 in 1994.

    Date of issuance: January 25, 1994
    Effective date: To be issued within 30 days from the date of 
issuance
    Amendment Nos.: 204, 204, and 201
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46228) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 25, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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

    Date of application for amendments: February 19, 1993
    Brief description of amendments: The amendments revise the Appendix 
A TSs 3.4.9.1, 3.4.9.2, and 4.4.9.2 relating to pressurizer surge line 
stratification.
    Date of issuance: January 31, 1994
    Effective date: January 31, 1994
    Amendment Nos.: 179 and 59
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: Aspril 28, 1993 (58 FR 
25854) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 31, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

    Date of application for amendment: February 24, 1993
    Brief description of amendment: The amendment revised the 
containment internal pressure lower limit of Technical Specification 
Figure 3.6-1 from 12.8 to 13.2 psia.
    Date of issuance: February 3, 1994
    Effective date: 30-days from date of issuance
    Amendment No.: 156
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 31, 1993 (58 FR 
16858) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 3, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of application for amendments: October 4, 1993
    Brief description of amendments: The amendments revise the 
surveillance test schedule in TS 4.6.1.2a and the associated Bases for 
performing Type A test which determine the overall integrated 
containment leakage rate.
    Date of issuance: January 11, 1994
    Effective date: January 11, 1994Amendment Nos. 158, 152Facility 
Operating Licenses Nos. DPR-31 and DPR-41: Amendments revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59748) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County, 
Texas

    Date of amendment request: December 6, 1993
    Brief description of amendments: The amendments modify Technical 
Specification 3.7.1.2 by extending the allowed outage time for the Unit 
1 Train D turbine-driven auxiliary feedwater pump from 72 hours to 168 
hours. This change is a one-time-only extension to accommodate an 
augmented test program for the turbine driven auxiliary feedwater pump 
during the restart of Unit 1 from the 1993 outage.
    Date of issuance: January 25, 1994
    Effective date: January 25, 1994, to be implemented within 10 days 
of issuance.
    Amendment No.: 58
    Facility Operating License No. NPF-76. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67848). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 25, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

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

    Date of application for amendment: November 4, 1993
    Brief description of amendment: The amendment revises Clinton 
Technical Specification 3/4.8.1.1, ``AC Sources - Operating,'' by 
relocating the surveillance requirement to inspect the diesel 
generators in accordance with the manufacturer's recommendations to the 
preventive maintenance program.
    Date of issuance: January 31, 1994
    Effective date: January 31, 1994
    Amendment No.: 87
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64610) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 31, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library District, 310 N. Quincy Street, Clinton, Illinois 61727.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
IllinoisDate of application for amendment: November 4, 1993

    Brief description of amendment: The licensee proposed modifying 
Technical Specification 3/4.8.2.1, ``DC Sources - Operating,'' by 
deleting the requirement that the plant be shut down to perform the 
required battery capacity or service testing. Following discussions 
with the licensee, the staff has modified the licensee's proposal and 
approved a one-time only change to permit replacement of the Division 
IV battery subsystem at power.
    Date of issuance: February 2, 1994
    Effective date: February 2, 1994
    Amendment No.: 88
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64610) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 2, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook, 
Nuclear Plant, Unit No. 2, Berrien County, Michigan

    Date of application for amendment: Aspril 16, 1993, as supplemented 
September 28 and December 3, 1993
    Brief description of amendment: The amendment revises Technical 
Specifications to allow certain tests normally designated as 18-month 
surveillances to be delayed until the next refueling outage scheduled 
to begin August 6, 1994. Extensions for four groups of surveillances 
(Groups 1, 2, 6, 11) were previously approved for Unit 2 in Amendment 
158 dated December 22, 1993. This amendment grants approval for the 
extensions requested for the remaining 12 groups of surveillances and 
completes the staff's review of the licensee's April 16, 1993 (as 
supplemented) application.
    Date of issuance: January 26, 1994
    Effective date: January 26, 1994
    Amendment No.: 159
    Facility Operating License No. DPR-74. Amendment revises the 
Technical Specifications. Dates of initial notice in Federal Register: 
Asugust 4, 1993 (58 FR 41505) and December 21, 1993 (58 FR 67850)The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated January 26, 1994. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.Northeast 
Nuclear Energy Company, et al., Docket No. 50-336, Millstone Nuclear 
Power Station, Unit No. 2, New London County, Connecticut
    Date of application for amendment: June 11, 1993, supplemented by 
letter dated November 15, 1993
    Brief description of amendment: The amendment revises the pressure/ 
temperature (P/T) limits for the reactor vessel. Specifically, Figure 
3.4-2, ``Millstone Unit 2 Reactor Coolant System Presure-Temperature 
Limitations for 12 Full Power Years,'' on page 3/4 4-19, is revised to 
reflect the change in the curves and the title change to ``Millstone 
Unit 2 Reactor Coolant System Pressure-Temperature Limitations for 20 
EFPY.''
    Date of issuance: January 27, 1994
    Effective date: As of the date of issuance to be implemented 
within30 days.
    Amendment No.: 170
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39054) The November 15, 1993, submittal provided 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 January 27, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

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

    Date of application for amendments: December 22, 1992, as 
supplemented July 19, 1993 (Reference LAR 92-08)
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2. Specifically, TS Section 3/4.3.2, ``Engineered Safety 
Features Actuation System Instrumentation,'' would be revised to change 
the second level undervoltage trip setpoint and allowable values. 
Technical Specification 3/4.8.1, ``A.C. Sources,'' would also be 
changed to revise the diesel generator (DG) steady state voltage 
surveillance requirements. The second level undervoltage relay TS 
setpoint and allowable values will be changed to maintain acceptable 
voltages at the 480 volt and 120 volt buses during sustained degraded 
voltage conditions. The DG steady state voltage surveillance 
requirements will be changed to ensure that the diesel generators 
provide adequate voltage when required to power the vital loads.
    Date of issuance: January 6, 1994
    Effective date: 60 days from date of issuance
    Amendment Nos.: 86 & 85
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 3, 1993 (58 FR 
7002) The July 19, 1993 submittal provided clarifying information and 
did not affect the initial Federal Register notice and proposed no 
significant hazards consideration. The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated January 6, 
1994No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendments: July 6, 1993, as supplemented 
December 29, 1993 (Reference LAR 93-03)
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) 3/4.3.2, ``Engineered Safety Features 
Actuation System Instrumentation,'' Table 4.3-2, ``Engineered Safety 
Features Actuation System Instrumentation Surveillance Requirements,'' 
for the Diablo Canyon Power Plant Unit Nos. 1 and 2 to relax the slave 
relay test frequency for slave relays K612A, K614B, K615A, and K615B 
from quarterly to once per 18 months during refueling or extended cold 
shutdowns. The affected slave relays cause isolation of the charging 
and letdown portions of the chemical and volume control system, and 
actuate charging pump suction valves associated with volume control 
tank and refueling water storage tank isolation.
    Date of issuance: January 31, 1994
    Effective date: For cycle 7 and after
    Amendment Nos.: 87 and 86
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR 
43929) The December 29, 1993, submittal provided clarifying information 
and did not effect the initial Federal Register Notice and proposed no 
significant hazards consideration.The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated January 31, 
1994.No significan 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.

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

    Date of application for amendments: January 9, 1991, as 
supplemented on August 19, 1991, June 22, 1992 and August 3, 1992
    Brief description of amendments: The amendment changed the 
Technical Specifications to revise the isolation setpoints for the 
ambient temperature switches for the High Pressure Coolant Injection 
and Reactor Core Isolation Cooling Systems room area coolers.
    Date of issuance: January 31, 1993
    Effective date: January 31, 1993
    Amendment Nos.: 132 and 99
    Facility Operating License Nos. NPF-14 and NPF-22. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 9, 1993 (58 FR 
32389) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 31, 1993.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Philadelphia Electric Company, Docket No. 50-353, Limerick 
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: Asugust 27, 1993, as 
supplemented November 10, and December 20, 1993
    Brief description of amendment: The amendment allows a one-time TS 
change to extend the allowed outage time (AOTs) for the Unit 2 residual 
heat removal service water (RHRSW) system as well as the suppression 
pool spray and suppression pool cooling modes of the residual heat 
removal system from 72, 168 (i.e. seven days), and 72 hours, 
respectively, to 288 hours (i.e., twelve days). The extended AOTs would 
allow continued Unit 2 operation while maintenance isolation valves are 
installed on both loops of the RHRSW system.
    Date of issuance: January 26, 1994
    Effective date: January 26, 1994
    Amendment No.  30
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50970) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 26, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Public Service Electric and Gas 
CompanyDelmarva 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 5, 1993
    Brief description of amendments: This amendment revised the Plant 
Operating Review Committee review, the Nuclear Review Board review, the 
Radiological Environmental Monitoring Program requirements, position 
titles, and the organization chart in Appendix B of the Technical 
Specifications (TS) to be consistent with Appendix A of the TS.
    Date of issuance: January 26, 1994
    Effective date: January 26, 1994Amendments Nos.: 183 and 188
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64612) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 26, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendment: February 2, 1993, and 
supplemented by letter dated November 16, 1993.
    Brief description of amendment: The amendment extends the period of 
time to reduce the setpoints of the Average Power Range Monitors and 
the Rod Block Monitor when the plant enters single-loop operations. 
Additionally, the change incorporates updated core values relative to 
single loop operations and the addition of a new Specification 3.0.5 
and its associated Bases.
    Date of issuance: January 25, 1994
    Effective date: As of date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment No.: 63
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 31, 1993 (58 FR 
16872) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 25, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: May 21, 1993, as supplemented on 
October 29, 1993, and November 16, 1993; the staff's proposed finding 
of no significant hazards is not affected by these supplements.
    Brief description of amendment: This amendment revises Technical 
Specifications surveillance requirement 4.4.2.2 to apply only to the 
pilot stage assembly of the safety relief valves (SRVs) and adds a new 
surveillance requirement which will require the main portion of the 
SRVs to be set pressure tested at least once every 5 years.
    Date of issuance: January 27, 1994
    Effective date: January 27, 1994
    Amendment No.: 64
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR 
43931) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 27, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: May 21, 1993 as supplemented on 
August 23, 1993.
    Brief description of amendment: This amendment revised a Technical 
Specification surveillance requirement to increase the voltage limit 
from 4580 to 4785 volts when performing the 18-month emergency diesel 
generator full load rejection test.
    Date of issuance: February 4, 1994
    Effective date: Effective as of date of issuance and to be 
implemented upon restart following fifth refueling outage currently 
scheduled to begin on March 5, 1994.
    Amendment No.: 65
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34091) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 4, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: Asugust 30, 1993
    Brief description of amendment: The amendment changes the main 
feedwater system containment isolation valves from the feedwater 
control and control bypass valves to the feedwater stop check valves.
    Date of issuance: January 21, 1994
    Effective date: As of date of issuance and shall be implemented 
within 60 days of the date of issuance
    Amendment No.  128
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR50974) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 21, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendments: November 25, 1991
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3/4.7.8, ``Fire Suppression Systems.'' This TS 
revision deletes the phrase ``during shutdown'' from the fire pump 
diesel engine surveillance requirement 4.7.8.1.2.c. This will allow the 
surveillance of the fire pump diesel engine to be performed when one or 
both Units 2 and 3 are in operation.
    Date of issuance: February 1, 1994
    Effective date: February 1, 1994
    Amendment Nos.: 109 and 98
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1992 (57 FR 
2600) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 1, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

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

    Date of application for amendments: September 30, 1993 (TS 345)
    Brief description of amendment: The amendment deletes conditions 
from the Browns Ferry Units 1, 2, and 3 licenses which require 
maintenance of positive access controls for the containment in 
accordance with 10 CFR 73.55(d)(8), and deletes a redundant condition 
from the Unit 3 license.
    Date of issuance: February 1, 1994
    Effective date: February 1, 1994
    Amendment Nos.: 202 - Unit 1; 221 - Unit 2; 175 - Unit 3
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendment revises the license conditions.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64616) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 1, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Asthens Public Library, South 
Street, Athens, Alabama 35611.

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

    Date of application for amendment: May 6, 1993
    Brief description of amendment: The amendment revises the reporting 
frequency requirements from semiannual to annual for submission to the 
NRC of the Radioactive Effluent Release Report, and clarifies the 
reporting requirements regarding steam generator tube inspection 
Category C-3 results.
    Date of issuance: December 30, 1993
    Effective date: December 30, 1993
    Amendment No.: 184
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34096) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 30, 1993.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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: July 16, 1993, as supplemented 
November 15, 1993. The November 15, 1993, submittal did not expand the 
scope of the original application and did not change the proposed no 
significant hazards consideration determination.
    Brief description of amendments: These amendments implement the 
revised 10 CFR Part 20, Standards for Protection Against Radiation, and 
reflect revisions to 10 CFR 50.36a.
    Date of issuance: January 25, 1994
    Effective date: January 25, 1994
    Amendment Nos. 185 and 185
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: Asugust 18, 1993 (58 FR 
43937) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 25, 1994.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: March 19, 1993, as supplemented 
December 9, 1993.
    Brief description of amendments: These amendments address plant 
operation with a control rod urgent alarm failure, a change in the 
control rod assembly partial movement surveillance test frequency, and 
proposed administrative changes.
    Date of issuance: February 4, 1994
    Effective date: February 4, 1994
    Amendment Nos. 186 and 186
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28064) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 4, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of application for amendments: January 14, 1993
    Brief description of amendments: The amendments split Technical 
Specification (TS) 15.3.1.E.2, which defines the allowable limits of 
chloride and fluoride in the reactor coolant, into two individual 
Limiting Conditions for Operation (LCOs), thus clarifying the reactor 
coolant chemistry limitations. In addition, the amendments added a 24-
hour hot shutdown action statement to the reactor coolant impurity 
limit LCOs. The amendments also modified the corresponding TS Bases 
Section.
    Date of issuance: January 27, 1994
    Effective date: January 27, 1994
    Amendment Nos.: 145 and 149
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12270) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 27, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By March 18, 1994, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC 20555 and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: November 18, 1993, as supplemented by 
letter dated December 21, 1993.
    Brief description of amendment: The amendment revises the River 
Bend, Unit 1 Technical Specifications to permit extending the time to 
perform leak rate testing of certain containment isolation valves and 
pressure isolation valves so that the testing can be performed during 
the refueling outage scheduled to start April 16, 1994, rather than 
requiring an earlier shutdown solely to perform the testing. Also, an 
exemption to 10 CFR Appendix J was issued on February 2, 1994, that 
provides an extension, consistent with the revision to the technical 
specifications, to allow the testing of containment isolation valves to 
be delayed until the refueling outage.
    Date of issuance: February 2, 1994
    Effective date: February 2, 1994
    Amendment No.: Asmendment No. 71
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. January 5, 1994 (59 FR 616)The 
Commission's related evaluation of the amendment, finding of emergency 
circumstances, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated February 2, 
1994.
    Attorney for licensee: Mark Wetterhahn, Esq., Bishop, Cook, Purcell 
and Reynolds, 1401 L Street, NW., Washington, D.C. 20005
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Pennsylvania Power and Light Company, Docket No. 50-388, 
Susquehanna Steam Electric Station, Unit 2, Luzerne County, 
Pennsylvania

    Date of application for amendment: January 24, 1994
    Brief description of amendment: The amendment revised the 
applicability requirement in Sections 3.0.4, 4.0.4, 3.3.7.5 Action 80, 
4.3.7.5, 3.4.2 Action c, and 4.4.2 of the Technical Specifications to 
permit Susquehanna, Unit 2 to continue to operate with the acoustic 
monitor on the ``S'' safety/relief valve tailpipe inoperable.

    Date of issuance: January 31, 1994
    Effective date: As of its date of issuance and will remain in 
effect until the next shutdown of sufficient duration to allow for 
containment entry, not to exceed the sixth refueling and inspection 
outage.
    Amendment No.: 100
    Facility Operating License No. NPF-22: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. On January 27, 1994, the staff 
issued a Notice of Enforcement Discretion, which was immediately 
effective and remained in effect until this amendment was issued.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, consultation with the Commonwealth of 
Pennsylvania and final no significant hazards considerations 
determination are contained in a Safety Evaluation dated January 31, 
1994.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge 2300 N Street NW., Washington, D.C. 20037
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18071.

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

    Date of application for amendment: January 13, 1994
    Brief description of amendment: The amendment modified the 
Technical Specifications (TS) to defer response time testing for low 
pressure emergency core cooling systems (ECCS) until startup following 
the next cold shutdown, but not later than the startup following 
completion of the spring 1994 refueling outage.
    Date of issuance: January 31, 1994
    Effective date: January 31, 1994
    Amendment No.: 120
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications. Public comments on proposed no significant 
hazards consideration comments received: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated January 31, 1994.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Project Director: The odore R. Quay
    Dated at Rockville, Maryland, this
    For the Nuclear Regulatory Commission
Robert A. Capra,
Acting Director, Division of Reactor Projects -I/II, Office of Nuclear 
Reactor Regulation
[Doc. 94-3465 Filed 2-15-94; 8:45 am]
BILLING CODE 7590-01-F