[Federal Register Volume 59, Number 81 (Thursday, April 28, 1994)]
[Unknown Section]
[Page ]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-10011]


[Federal Register: April 28, 1994]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving no Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 2, 1994, through April 15, 1994. The 
last biweekly notice was published on April 13, 1994 (59 FR 17591).

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 May 27, 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.

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

    Date of Amendment Requests: January 4, 1994.
    Description of Amendment Requests: The proposed amendment would 
change Technical Specification 3/4.2.3 Azimuthal Power Tilt and its 
associated bases. The licensee proposed to change the Azimuthal Power 
Tilt limit from less than or equal to 10 percent to less than or equal 
to 3 percent when the Core Operating Limit Supervisory System is out of 
service.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:

    Standard 1--Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Decreasing the COLSS [Core Operating Limit Supervisory System] 
out-of service Azimuthal Power Tilt Technical Specification limit 
does not increase the probability or consequences of an accident 
previously evaluated. The Technical Specification operating limit is 
being conservatively reduced to conform to the assumptions used in 
the safety analysis. The reduced operating limit requires a more 
uniform power distribution in the reactor core. The uniform power 
distribution may reduce the consequences of an accident previously 
evaluated by not allowing regions in the core to operate at higher 
power levels.
    Standard 2--Create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed amendment will result in an alarm setpoint change, 
but does not involve any equipment changes and will not alter the 
manner in which the plant will be operated. For this reason, this 
amendment will not create the possibility of an new or different 
kind of accident from any previously evaluated. The proposed 
operating range is smaller and completely within the existing 
Technical Specification limits; thus, there are no mechanisms to 
create the possibility of a new or different kind of accident from 
those previously evaluated.
    Standard 3--Involve a significant reduction in a margin of 
safety.
    The proposed amendment conservatively reduces the COLSS out-of-
service Azimuthal Power Tilt Technical Specification limit, thereby 
increasing the margin of safety. The proposed operating range is 
smaller and completely bounded by the existing Technical 
Specification limits.

    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room Location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.
    Attorney for Licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: Theodore R. Quay.

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

    Date of Amendment Request: March 25, 1994.
    Description of Amendment Request: The amendment would revise 
Technical Specification 3/4.8.4.2, Motor Operated Valves Thermal 
Overload Protection, with a more accurate description of the motor-
operated valve (MOV) bypass configuration.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change is administrative in nature, providing a more 
accurate description of the MOV electrical supply configuration 
related to the thermal overload bypass function. Therefore, the 
change in terminology would not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve any modifications or 
additions to plant equipment and the design and operation of the 
plant will not be affected. Therefore, the change in MOV thermal 
overload bypass function terminology would not increase the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed terminology change does not affect any parameters 
which relate to the margin of safety as defined in the Technical 
Specifications or in the FSAR [Final Safety Analysis Report]. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for Licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman.

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

    Date of Application for Amendments: March 7, 1994, as supplemented 
on March 24, 1994.
    Description of Amendment Requests: The proposed amendment would 
change Technical Specification 4.6.1.2 by removing the specific 
schedules for containment integrated leak rate testing (CILRT) and 
specifying that the testing will be done in accordance with Appendix J 
to 10 CFR part 50.
    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:

    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change will allow flexibility in the scheduling for 
Type A tests in the 10-year service period while still meeting the 
requirements in 10 CFR 50 Appendix J. Additional flexibility is 
needed for plants using an 18-month fuel cycle to allow refueling 
outages and 10-year inservice testing intervals to coincide. For 
performance of the third Type A test at Byron, the change would 
allow an extension of four (4) months beyond the current maximum 50-
month surveillance interval. The third test would be completed at 
the fifty-four (54) month interval for Byron Units 1 and 2.
    For Braidwood Units 1 and 2, an extension on the surveillance 
time interval will not be necessary to satisfy the requirements of 
Appendix J. The Braidwood Units have scheduled the third Type A test 
to be conducted with the 10-year Inservice Inspection.
    The results of the previous Type A leak tests show the overall 
leakage from the Byron containment buildings at very low levels. The 
extension of the Type A test by four months would not cause the 
consequences of a previously evaluated accident to increase. By 
continuing to conform to the requirements of 10 CFR 50 Appendix J, 
the test frequency, methodology, and acceptance criteria for 
containment leakage remains the same. Therefore, there is no 
significant increase in the probability or the consequences of an 
accident previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not affect the design or operation of 
any system, structure or component in the plant. There are no 
changes to parameters governing plant operation and no new or 
different type of equipment will be installed. No new accident 
scenarios are created by the proposed change because the test 
frequency continues to meet the requirements of Appendix J of 10 CFR 
part 50. There is no affect on containment structure, the 
penetrations, or the facility. The proposed change to the test 
schedule only provides flexibility in meeting the same requirement 
for three tests in a 10-year period. The testing method and bases 
have not changed. Therefore, operation of the units with this more 
flexible test schedule will not result in an accident previously not 
analyzed in the Updated Final Safety Analysis Report (UFSAR). The 
proposed changes do not impact the design bases of the containment 
and do not modify the response of the containment during a design 
basis accident. Therefore, the changes do not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the margin of safety for any 
Technical Specifications. The initial conditions and methodologies 
used in the accident analyses remain unchanged, therefore, the 
results of the accident analyses are not impacted. The proposed 
change to the schedule allows for additional flexibility in meeting 
the requirement for three tests in a 10-year period. Elimination of 
the specified time interval for Type A testing would allow Byron 
Units 1 and 2 to extend the surveillance requirement of the third 
Type A test by four (4) months. This would exceed the existing 
maximum 50 month interval currently specified in Technical 
Specifications. The extension will allow performance of the Type A 
test to coincide with the seventh refueling outage, 10 year 
Inservice Inspection, and continue to meet the requirements of 
Appendix J to 10 CFR part 50. These proposed changes do not affect 
or change any limiting conditions for operation (LCO), or any other 
surveillance requirements in the Technical Specifications.
    The results of the previous Type A leak tests have shown that 
the overall leakage rates from the Byron containment buildings were 
at low levels. The latest test results for Units 1 and 2 were 0.0175 
weight percent per day and 0.0376 weight percent per day, 
respectively. The overall containment leakage rates have 
consistently remained well below the acceptance criteria for Byron 
Station Type A tests of 0.075 weight percent per day. The testing 
method, acceptance criteria, and bases for the surveillance 
requirement will not be changed by the proposed amendment.
    The present test performance margins, coupled with the Type B & 
C test program for monitoring and repairing individual leakage 
components provides justification for the proposed change. The Type 
B & C tests provide added assurance that the overall containment 
integrated leakage rates remain satisfactory. No significant leakage 
trends have been identified which threaten the overall containment 
leakage specifications.
    In summary, Commonwealth Edison concludes that this change does 
not involve a significant reduction in a margin of safety because 
the containment integritiy will be maintained. Testing in accordance 
with Appendix J requirements ensures confidence is containment 
intergity. The proposed Technical Specifications amendment will 
continue to require testing that is consistent with Appendix J 
requirements. Additionally, results from previous tests have shown 
acceptable low overall containment leakage rates. Extension of Type 
A testing for four months would not involve a signficant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481
    Attorney for Licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    NRC Project Director: James E. Dyer.

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: February 18, 1994

    Description of Amendment Request: This amendment is an additional 
followup to the amendment request of May 29, 1992, published in the 
Federal Register on July 8, 1992 (57 FR 30242), which changed the 
Technical Specifications Sec. 1.0, Definitions, to accommodate a 24-
month fuel cycle and which proposed the extension of the test intervals 
for specific surveillance tests. This amendment proposes extending the 
surveillance intervals to 24 months for the following additional 
surveillance tests:

    (1) Analog Rod Position Indication.
    (2) Plant Noble Gas Activity Monitor (R-44).
    (3) Low Turbine Auto Stop Oil Pressure Reactor Trip.
    (4) 6.9 KV Undervoltage Relays.
    (5) Boric Acid Tank Level.
    (6) Vapor Containment Sump Discharge Flow and Temperature 
Channel.
    (7) Loss of Power Undervoltage and Degraded Voltage Relays.
    (8) Over-pressurization Protection System (OPS) and Control Rod 
Protection System (for use with Low Parasite [LOPAR] fuel) Trip.
    (9) Condenser Evacuation System Activity Monitor (R-45).
    (10) Service Water Inlet Temperature Monitoring Instrumentation.
    (11) Sampler Flow Rate Monitors.
    (12) Boric Acid Makeup Flow System.
    (13) Plant Vent Noble Gas Effluent Monitor (R-27).

    The amendment also proposes to change the surveillance interval for 
the Refueling Water Storage Tank Level to quarterly and to change the 
trip setpoint for the Control Rod Protection System. The changes 
requested by the licensee are in accordance with Generic Letter 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle.''
    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) Analog Rod Position Indication:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
analog rod position indication channel be changed from every 18 
months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that none of the major error contributors are time 
dependent and that it can be reasonably expected that the channel 
will remain within calibration tolerance over a possible 30 month 
operating cycle. In addition, the rod bottom bistable is subject to 
monthly testing which would detect any abnormalities in an extended 
operating cycle. Due to this monthly test and the acceptable past 
test history, it is concluded that the channel will continue to 
operate within tolerance over an extended operating cycle and will 
not contribute to 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 accident previously evaluated has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval is not expected to affect the 
ability of the instrument channel to remain within calibration 
tolerance. Furthermore, the rod position indicator is used in normal 
operation only as an aid in control rod movement. Normally, very 
little control rod movement occurs during normal operation. 
Furthermore, it is not relied upon for accident prevention or 
accident mitigation. In accordance with existing Technical 
Specifications, normal operation can continue even if one channel is 
inoperable because alternate means (core instrumentation) exists to 
monitor rod position. The frequent monthly test tends to minimize 
the effect of a longer operating cycle for the rod position 
indication channel as any malfunction induced by time would be 
detected. Thus, it is concluded that the possibility of a new or 
different kind of accident from any accident previously evaluated 
has not been created.
    3. A significant reduction in a margin of safety is not 
involved.
    A statistical analysis of past calibration data has not 
identified any time dependent error contributors. Also, past test 
data indicates that the channel remains within calibration tolerance 
over the existing operating cycle. A longer operating cycle would 
increase the risk of drift, however accuracy is not a prime 
requirement for the RPI. Therefore, it is concluded that a longer 
operating cycle will not result in a significant reduction in a 
margin of safety.
    [(2) Plant Noble Gas Activity Monitor (R-44):]
    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.
    It is proposed that the channel calibration frequency for the 
Plant Noble Gas Activity Monitor (R-44) be changed from every 18 
months (+25%) to every 24 months (+25%).
    The function of R-44 is to respond to high activity levels 
during normal operation.
    The setpoint for R-44 is established sufficiently above the 
expected radioactivity level in the discharge stream to preclude 
false actions but sufficiently below the allowed discharge 
radioactivity concentration so that discharge in excess of 
permissible limits does not occur. Monitor readouts are not used for 
quantitative purposes, but are used to respond to relative changes 
in radioactivity concentration.
    There is limited data to support an unqualified extension of the 
surveillance interval. However, the instrument is checked for 
operability prior to release. Should the instrument be inoperable 
releases may continue provided grab sample analysis is performed. 
Since the monitor is subject to daily channel checks, monthly source 
checks, and quarterly functional channel tests, abnormal instrument 
behavior or inoperability would be detected permitting corrective 
actions during the extended surveillance interval.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Operability of the instrument is important rather than ability 
to maintain a specific setpoint. Operability of the instrument is 
verified prior to a planned discharge and this is independent of an 
extended surveillance cycle.
    3. There has been no reduction in the margin of safety.
    As the Technical Specifications permit pre-planned release even 
with an inoperable instrument, the margin of safety is not impacted 
by an extended surveillance interval provided that instrument 
operability is verified prior to release. This is also required by 
the Technical Specifications.
    [(3) Low Turbine Auto Stop Oil Pressure Reactor Trip:]
    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.
    It is proposed that the channel calibration frequency for the 
Low Turbine Auto Stop Oil Pressure system be changed from every 18 
months (+25%) to every 24 months (+25%).
    No credit is taken for a reactor trip from a low turbine auto 
stop oil pressure signal resulting from a turbine trip. Rather, the 
safety analysis assumes this reactor trip does not occur during full 
load rejection until an overpower delta T condition causes a reactor 
trip. In addition, no credit is taken for this system for turbine 
missile protection. Therefore, extending the surveillance interval 
for this parameter has no impact upon the probability or 
consequences of an accident.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    As no credit is taken in the safety analysis for this trip, the 
possibility of a new or different kind of accident has not been 
created by extending the surveillance interval.
    3. There has been no reduction in the margin of safety.
    Past test results have not identified any failures. Therefore, 
pursuant to Generic Letter 91-04, it is reasonably expected that 
this system will continue to function in an acceptable manner over 
an extended operating cycle.
    [(4) 6.9 kv Undervoltage Relays:]
    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.
    It is proposed that the calibration frequency for the 6.9 kv 
undervoltage channel be changed from every 18 months (+25%) to every 
24 months (+25%).
    Quarterly testing of these relays is required by Technical 
Specifications. The data from the quarterly tests of the new relays 
will be used to assure that drift does not exceed projected values. 
The quarterly tests provide a means of maintaining calibration 
within specified values, virtually eliminating any impact upon 
safety from an extended operating cycle.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Because the quarterly tests assure that relay performance 
remains within specified limits, there is no possibility of creating 
a new or different kind of accident from any previously analyzed.
    3. There has been no reduction in the margin of safety.
    The requirement for a channel functional test each quarter 
minimizes any potential impact upon safety due to an extended 
operating cycle.
    [(5) Boric Acid Tank Level:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequence of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
Boric Acid Tank Level instrumentation be changed from every 18 
months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists between the existing 
Technical Specification limit and the licensing basis Safety 
Analysis limit to accommodate the channel statistical error 
resulting from a 30 month operating cycle. The existing margin 
between the Technical Specification limit and the Safety Analysis 
limit provides assurance that plant protective actions will occur as 
required. It is therefore concluded that changing the surveillance 
interval from 18 months (+25%) to 24 months (+25%) will not result 
in 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 accident previously evaluated has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin between the 
existing Technical Specification limit and the Safety Analysis 
limit. Plant equipment, which will be set at (or more conservatively 
than) Technical Specification limits, will provide protective 
functions to assure that Safety Analysis limits are not exceeded. 
This will prevent the possibility of a new or different kind of 
accident from any previously evaluated from occurring.
    3. A significant reduction in a margin of safety is not 
involved.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the margin which exists between the current 
Technical Specification limit and the licensing basis Safety 
Analysis limit. This margin, which is equivalent to the existing 
margin, is necessary to assure that protective safety functions will 
occur so that Safety Analysis limits are not exceeded.
    [(6) Vapor Containment Sump Discharge Flow and Temperature 
Channel:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the calibration frequency for the VC sump 
discharge flow and temperature channel be changed from every 18 
months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists between the existing 
Technical Specification and the licensing basis Safety Analysis to 
accommodate the channel statistical error resulting from a 30 month 
operating cycle. The existing margin between the Technical 
Specification and the Safety Analysis provides assurance that plant 
protective actions will occur as required. It is therefore concluded 
that changing the surveillance interval from 18 months (+25%) to 24 
months (+25%) will not result in 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 accident previously evaluated has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin between the 
existing Technical Specification and the Safety Analysis. Plant 
equipment, which will be set at (or more conservatively than) 
Technical Specification limits, will provide protective functions to 
assure that Safety Analysis limits are not exceeded. This will 
prevent the possibility of a new or different kind of accident from 
any previously evaluated from occurring.
    3. A significant reduction in a margin of safety is not 
involved.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the margin which exists between the current 
Technical Specification and the licensing basis Safety Analysis. 
This margin, which is equivalent to the existing margin, is 
necessary to assure that protective safety functions will occur so 
that Safety Analysis limits are not exceeded.
    [(7) Loss of Power Undervoltage and Degraded Voltage Relays:]
    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.
    The Technical Specifications specify that the Loss of Power 
(undervoltage and degraded voltage) relays be calibrated and tested 
at a refueling interval; that the undervoltage alarm be calibrated 
at a refueling interval, and that the undervoltage (station 
blackout) input to Auxiliary Feedwater be calibrated at refueling 
intervals. It is proposed that the surveillance frequency be revised 
from 18 months (+25%) to 24 months (+25%).
    All of the undervoltage and station blackout relays were found 
to be within specification at each of the refueling outage 
calibration periods.
    Since the old relays have been replaced with relays from a 
different manufacturer whose drift characteristics are expected to 
be superior, extending the surveillance interval by several months 
will not significantly increase the probability or consequences of 
an accident.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Past test results provide reasonable assurance that the relays 
will perform in an acceptable manner for an extended operating 
cycle. With the installation of the new relays, whose performance 
will surpass the old relays, it is concluded that the plant will 
perform within its design basis for an extended operating cycle. 
Therefore, the possibility of a new or different kind of accident 
from any previously analyzed has not been created.
    3. There has been no significant reduction in the margin of 
safety.
    Since the new relays will surpass the performance of the old 
relays, there is reasonable assurance that a significant reduction 
in the margin of safety has not resulted from an extended operating 
cycle.
    [(8) Over-pressurization Protection System (OPS) and Control Rod 
Protection System (for use with Low Parasite (LOPAR) fuel) Trip:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
Over-pressurization protection system and the LOPAR trip system be 
changed from every 18 months (+25%) to every 24 months (+25%). This 
necessitates a change in the LOPAR Technical Specification trip 
setpoint from 350  deg.F to 381  deg.F.
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed based upon historical test data. 
Based on this analysis, a change to the Technical Specifications is 
required. Sufficient margin exists between the Safety Analysis limit 
and the proposed Technical Specification limit to accommodate 
projected channel uncertainty over a 30 month operating cycle. A 
statistical basis exists to assure that protective action will occur 
to prevent Safety Analysis limits from being exceeded. Thus, there 
will not be a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident 
previously evaluated has not been created.
    Based upon a statistical analysis of past historical test data 
it has been demonstrated that reasonable assurance exists to 
conclude that Safety Analysis limits will not be exceeded over a 30 
month operating cycle. The proposed Technical Specification limits 
provide margin with respect to the Safety Analysis limits and 
confidence that appropriate plant protective response will be 
provided to prevent the possibility of a new or different kind of 
accident from that previously evaluated from being created.
    3. A significant reduction in a margin of safety is not 
involved.
    The proposed changes to the Technical Specification limits are 
being made to assure that the previously established margin remains 
the same between plant protective function set points and Safety 
Analysis limits. This margin is based upon an evaluation of past 
historical test data and analytical methods for projecting 
instrument channel uncertainty over a 30 month operating cycle. It 
is therefore concluded that the existing margin of safety has been 
preserved.
    [(9) Condenser Evacuation System Activity Monitor (R-45):]
    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.
    It is proposed that the channel calibration frequency for the 
Condenser Evacuation System Noble Gas Activity Monitor (R-45) be 
changed from every 18 months (+25%) to every 24 months (+25%).
    Since this radiation monitor is relatively new a degree of 
uncertainty is introduced by extending the surveillance interval by 
several months. However, the setpoint for automatic diversion is set 
some what conservatively. It is established sufficiently high to 
avoid spurious actuations and yet sufficiently low so that diversion 
and alarm can occur should a step increase in radioactivity level 
occur. Under these circumstances considerable departure from the 
setpoint can be accommodated and the monitor will still perform its 
intended safety function. Continued monitor operability is important 
and malfunction would be detected by monthly checks during the 
extended operating cycle. Thus, despite the introduction of a new 
monitor, the capability of R-45 to tolerate drift in addition to 
monthly operator checks, leads to the conclusion that an extended 
operating cycle will not result in a significant increase in the 
probability or consequences of an accident.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Monthly checks would identify abnormal operating 
characteristics, should the instrument fail to perform its intended 
function. In the event of tube rupture with a reactor coolant system 
radioactivity concentration corresponding to 1% defective fuel, the 
resultant site boundary dose would be within 10 CFR [part] 20 limits 
should the monitor fail to perform its function (as discussed in 
FSAR [Final Safety Analysis Report]). In addition, alternate means 
of alarms to indicate a tube rupture event are available. Thus, the 
possibility of a new or different kind of accident has not been 
created.
    3. There has been no reduction in the margin of safety.
    Although this monitor is not necessary to mitigate releases 
below regulatory limits, it does provide the earliest of a steam 
generator tube leak. In this regard, continued instrument operation 
is important. Continued instrument operability would be verified by 
the monthly checks in an extended operating cycle.
    [(10) Service Water Inlet Temperature Monitoring 
Instrumentation:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
Service Water Inlet Temperature Monitoring Instrumentation be 
changed from every 18 months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists between the existing 
Technical Specification and the licensing basis Safety Analysis to 
accommodate the channel statistical error resulting from a 30 month 
operating cycle. The existing margin between the Technical 
Specification and the Safety Analysis provides assurance that plant 
protective actions will occur as required. It is therefore concluded 
that changing the surveillance interval from 18 months (+25%) to 24 
months (+25%) will not result in 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 accident previously evaluated has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin between the 
existing Technical Specification and the Safety Analysis. Plant 
equipment, which will be set at (or more conservatively than) 
Technical Specification limits, will provide protective functions to 
assure that Safety Analyses are not exceeded. This will prevent the 
possibility of a new or different kind of accident from any 
previously evaluated from occurring.
    3. A significant reduction in a margin of safety is not 
involved.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the existing margin between the current 
Technical Specification and the licensing basis Safety Analysis. 
This margin, which is equivalent to the existing margin, is 
necessary to assure that the protective safety functions occur and 
that the Safety Analysis limits are not exceeded.
    [(11) Sampler Flow Rate Monitor:]
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the channel calibration frequency for the 
Sample Flow Rate Monitors be changed from every 18 months (+25%) to 
every 24 months (+25%).
    The flow rate monitors are used to estimate the total volume of 
air passed through filters. There is no setpoint or safety function 
served by these monitors. A high level of radioactivity in the 
discharge stream is detected by R-43 and/or R-44.
    Insofar as discharge via the unit vent is permissible with the 
monitors inoperable, extension of the surveillance interval will 
have no impact upon safety.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    As the nuclear safety function is provided by other monitors in 
the event of high radioactivity levels in the discharge stream, 
extension of the surveillance interval will have no impact upon the 
creation of a new or different kind of accident.
    3. There has been no reduction in the margin of safety.
    These flow monitors are utilized to determine the total air flow 
through filters for computational purposes. As adequate measures 
(other monitors) exist to prevent the possibility of discharging 
radioactivity in excess of applicable limits, there is virtually no 
impact upon safety incurred by extending the surveillance interval.
    [(12) Boric Acid Makeup Flow System:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
Boric Acid Makeup Flow System be revised from every 18 months (+25%) 
to every 24 months (+25%). A statistical analysis of channel 
uncertainty for a 30 month operating cycle has been performed. Based 
upon this analysis it has been concluded that sufficient margin 
exists between the existing Technical Specification limit and the 
licensing basis Safety Analysis limit to accommodate the channel 
statistical error resulting from a 30 month operating cycle. The 
existing margin between the Technical Specification limit and the 
Safety Analysis limit provides assurance that plant protective 
actions will occur as required. It is therefore concluded that 
changing the surveillance interval from 18 months (+25%) to 24 
months (+25%) will not result in 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 accident previously evaluated has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin between the 
existing Technical Specification limit and the Safety Analysis 
limit. Plant equipment, which will be set at (or more conservatively 
than) Technical Specification limits, will provide protective 
functions to assure that Safety Analysis limits are not exceeded. 
This will prevent the possibility of a new or different kind of 
accident from any previously evaluated from occurring.
    3. A significant reduction in a margin of safety is not 
involved.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the margin which exists between the current 
Technical Specification limit and the licensing basis Safety 
Analysis limit. This margin, which is equivalent to the existing 
margin, is necessary to assure that protective safety functions will 
occur so that Safety Analysis limits are not exceeded.
    [(13) Plant Vent Noble Gas Effluent Monitor (R-27):]
    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.
    It is proposed that the channel calibration frequency for the 
Plant Vent Noble Gas Effluent Monitor (R-27) be changed from every 
18 months (+25%) to every 24 months (+25%).
    R-27 is a high range noble gas monitor intended for use after an 
accident to provide information about the magnitude of radioactive 
releases. It serves no purpose during normal operation. It provides 
no function to prevent or mitigate an accident but does provide a 
role in assessing the consequences of an accident. As the monitor is 
a high range monitor, an estimate of the magnitude of release rather 
than accuracy is important. Accordingly, continued operability of 
the instrument during an extended operating cycle is more important 
than the device exhibiting minimal drift characteristics. 
Malfunction of the instrument would be detected by the shift checks 
and functional tests performed during the extended operating cycle.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Since the monitor provides no preventive or mitigating action in 
the event of an accident, no new or different type of accident has 
been created by extending the operating cycle. In terms of post 
accident assessment capability, alternate means exist to assess 
offsite releases in the event of failure of this instrument.
    3. There has been no reduction in the margin of safety.
    Since the instrument provides no safety function and alternate 
means exist for post accident assessment purposes, there will be no 
impact on safety due to an extended period between calibrations.
    [(14) Refueling Water Storage Tank Level:]
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
RWST instrumentation be changed from every 18 months (+25%) to 
quarterly (once every 3 months).
    A statistical analysis of channel uncertainty for a 3 month 
surveillance has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists between the existing 
Technical Specification limit and the licensing basis Safety 
Analysis limit to accommodate the channel statistical error 
resulting from a 3 month quarterly surveillance. The existing margin 
between the Technical Specification limit and the Safety Analysis 
limit provides assurance that plant protective actions will occur as 
required. It is therefore concluded that changing the surveillance 
interval from 18 months (+25%) to quarterly will not result in 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 accident previously evaluated has not been created.
    The proposed change in surveillance interval will result in a 
channel statistical allowance which provides the necessary margin 
between the existing Technical Specification limit and the Safety 
Analysis limit. Plant equipment, which will be set at (or more 
conservatively than) Technical Specification limits, will provide 
protective functions to assure that Safety Analysis limits are not 
exceeded. This will prevent the possibility of a new or different 
kind of accident from any previously evaluated from occurring.
    3. A significant reduction in a margin of safety is not 
involved.
    The above change in surveillance interval will result in a 
channel statistical allowance which is necessary between the current 
Technical Specification limit and the licensing basis Safety 
Analysis limit. This margin is necessary to assure that protective 
safety functions will occur so that Safety Analysis limits are not 
exceeded.


    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: March 24, 1994.
    Description of Amendment Request: The changes are in support of the 
forthcoming Cycle 7 for Catawba, Unit 2. The proposed Technical 
Specification (TS) changes reflect:
    (1) An increase from 2000 parts per million (ppm) to 2175 ppm in 
the required spent fuel storage pool minimum boron concentration during 
Modes 1-3 operation,
    (2) An increase from 2000 ppm to 2175 ppm in the required reactor 
coolant system (RCS) and refueling canal minimum boron concentration 
during Mode 6 operation,
    (3) The inclusion of two reload related topical reports into TS 
6.9.1.9, and
    (4) The revision of an administrative nature to correct errors in 
nomenclature and to remove obsolete footnotes.
    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:

Increase in Boron Concentration Limit for the Spent Fuel Storage Pool 
(Standby Makeup Pump Water Supply)

    The required spent fuel storage pool minimum boron concentration 
was increased from 2000 ppm to 2175 ppm during Modes 1-3.
    The proposed revision is conservative, and is required only to 
maintain consistency between the boron concentration of the spent 
fuel storage pool and the boron concentration of the RWST [refueling 
water storage tank] during Modes 1-3 operation. Therefore, there 
will be no adverse impact upon the probability or consequences of 
any previously analyzed accident.
    Likewise, the proposed change will not create the possibility of 
a new or different kind of accident, since no new failure modes are 
identified.
    Finally, no negative impact upon any safety margin is created 
since the proposed change is conservative.

Increase in Boron Concentration Limits for the RCS and Refueling Canal 
in Mode 6

    The increase in the required RCS and refueling canal minimum 
boron concentration was added only to maintain consistency between 
the boron concentration of the RCS and refueling canal and the RWST 
in Mode 6.
    The change in boron concentration limits for the RCS and 
refueling canal will not increase the probability of an accident 
since no accident initiators are involved with this change. Since 
the change is conservative, the consequences of an accident 
previously evaluated will not be increased. The increase in the 
boron concentration limit for the RCS and refueling canal in Mode 6 
adds further margin to the initial conditions assumed for the boron 
dilution accident in the safety analysis. Therefore, the 
consequences of the boron dilution accident previously evaluated 
will not be increased.
    The possibility of a new or different kind of accident from any 
previously evaluated will not be created since this change is 
bounded by previously evaluated accidents and does not introduce any 
new failure modes.
    This change does not involve a significant reduction in the 
margin of safety since the analyses performed demonstrate that the 
limits imposed meet all accident analysis and design basis 
requirements.

Addition of Two Reload Related Topical Reports

    This change is administrative in nature and adds two previously 
approved topical reports to the list of methodologies used to 
determine core operating limits. The change will have no impact upon 
either the probability or consequences of a previously analyzed 
accident. The methodologies described in the topical reports have 
been previously reviewed and approved by the NRC. Also, no new 
accident possibilities are created, since this is an administrative 
change. Finally, no impact upon any safety margin is created, since 
the change is administrative in nature and the described topical 
reports have received full NRC approval.

Correction of Errors in Nomenclature and Removal of Obsolete Footnotes

    These changes are also administrative in nature and are intended 
to correct miscellaneous errors and obsolete references. As such, 
the changes will have no impact upon either the probability or 
consequences of any previously analyzed accidents, will not create 
the possibility of any new accident scenarios, and will not impact 
any safety margins.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: 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: David B. Matthews.

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: March 23, 1994.
    Description of Amendment Request: The proposed amendments would 
revise Technical Specification (TS) 6.9.2, ``Core Operating Limits 
Report,'' to include a reference to a Duke Power Company (DPC) Topical 
Report describing an analytical method for determining the core 
operating limits.
    Specifically, the amendments would add: ``(4) DPC-NE-1004A, Nuclear 
Design Methodology Using CASMO-3/SIMULATE-3P, November 1992,'' to TS 
6.9.2.
    Basis for Proposed No Significant Hazards Consideration 
Determination: The NRC staff reviewed Topical Report DPC-NE-1004A and 
concluded in a Safety Evaluation Report dated November 23, 1992, that 
the described nuclear design methodology is acceptable for performing 
reload analyses for the DPC B&W 177-assembly cores in the Oconee units. 
The addition of this approved nuclear design methodology to those 
referenced in TS 6.9.2 provides an alternative method for determining 
core operating limits such that all applicable limits (e.g., fuel 
thermal mechanical limits, core thermal hydraulic limits, ECCS limits, 
nuclear limits such as shutdown margin, and transient and accident 
analysis limits) of the safety analysis are met. Therefore, the 
proposed change to the TS (1) does not involve a significant increase 
in the probability or consequences of an accident previously evaluated, 
(2) does not create the possibility of a new or different kind of 
accident than previously evaluated, and (3) does not involve a 
significant reduction in the margin of safety.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:
    Duke Power Company (Duke) has made the determination that this 
amendment request involves a No Significant Hazards Consideration by 
applying the standards established in 10 CFR 50.92. This ensures that 
operation of the facility in accordance with the proposed amendment 
would not:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    Each accident analysis addressed within the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to this 
amendment request. The Technical Specifications will continue to 
require operation within the bounds of the cycle-specific parameter 
limits. The cycle-specific parameter limits will be calculated using 
NRC approved methodology. The proposed amendment is simply an 
administrative change to update the list of NRC approved methods in 
Technical Specification 6.9.2. Therefore, the probability of any 
Design Basis Accident (DBA) is not affected by this change, nor are 
the consequences of a DBA affected by this change. This is because 
the addition of an NRC approved reference to Technical Specification 
6.9.2 is not considered to be an initiator or contributor to any 
accident analysis addressed in the Oconee FSAR.
    (2) Create the possibility of a new or different kind of 
accident from any kind previously evaluated:
    Operation of ONS [Oconee Nuclear Station] in accordance with 
these Technical Specifications will not create any failure modes not 
bounded by previously evaluated accidents. Consequently, this change 
will not create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety:
    The Technical Specifications will continue to require operation 
within the bounds of the cycle-specific parameter limits. Duke will 
continue to calculate the cycle-specific parameter limits using NRC 
approved methodology. In addition, each future reload will require a 
10 CFR 50.59 safety review to ensure that operation of the unit 
within the cycle-specific limits will not involve a reduction in a 
margin of safety. Therefore, no margins of safety are affected by 
the addition of an NRC approved methodology to Technical 
Specification 6.9.2.

    Based on the staff's analysis and its review of the licensee's 
analysis, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.
    Attorney for Licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036.
    NRC Project Director: David B. Matthews.

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: March 18, 1994.
    Description of Amendment Request: The proposed amendments would 
revise Technical Specification (TS) 3/4.3.3.6, Accident Monitoring 
Instrumentation, TS 3/4.6.4.1, Hydrogen Monitors, and their associated 
bases to incorporate the technical substance of Specification 3.3.3 
from NUREG-1431, Revision O (Standard Technical Specifications) for the 
Westinghouse Owners Group.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes affect instrumentation that would be used to 
assess the condition of the plant during and following an accident. 
As such, the changes can have no effect on the probability of any 
accident previously evaluated since this instrumentation has no 
bearing on initiating events. The proposed changes will continue to 
ensure the capability to monitor plant conditions during and 
following an accident by requiring redundancy or diversity and 
timely corrective action in the event of inoperable instrumentation. 
Therefore, the proposed changes will not significantly increase the 
consequences of any accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes affect the operability and action 
requirements for the post accident monitoring instrumentation 
system. Accordingly, the proposed changes do 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 nor has any new limiting failure been identified as a 
result of the proposed changes. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. The intent of the existing TS requirements is 
to ensure the capability to monitor the plant condition during and 
following an accident so that the operators will have the 
information necessary to monitor and evaluate the course of the 
event and take any necessary action. Under the proposed changes this 
capability will be maintained by ensuring redundancy or diversity 
and by requiring timely corrective action in the event of inoperable 
instrumentation. In addition, the proposed changes would avoid 
unnecessary plant shutdowns by specifying an appropriate level of 
action in response to inoperable instrumentation. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: 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: David B. Matthews.

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

    Date of Amendment Request: April 6, 1994.
    Description of Amendment Request: The proposed amendment changes 
the Technical Specifications to eliminate the main steam line radiation 
monitor(s) (MSLRMs) reactor scram and isolation functions of the MSLRMs 
currently contained in Tables 3.1.-1 and 4.1-1 of the Technical 
Specifications and the associated Bases statements. This action follows 
the recommendations of the BWR Owners Group (BWROG) in their Safety 
Evaluation, NEDO-31400A, previously approved by the NRC Staff on May 
15, 1991 by letter to the BWROG. Following is a brief description of 
the proposed changes:
    Tech. Spec. 3.1, ``Protective Instrumentation'' Bases is revised to 
delete reference to the paragraph describing the Main Steam Line (MSL) 
radiation monitoring functions for indication of excessive fuel failure 
and initiation of a reactor scram and MSL isolation.
    Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements 
- A. Reactor Scram Functions,'' is revised to delete line Item No. 7 - 
``High Radiation in Main Steam Line Tunnel.''
    Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements 
- B. Reactor Isolation Functions,'' is revised to delete line Item No. 
6 - ``High Radiation in Main Steam Line Tunnel.''
    Tech. Spec. Table 3.1.1., ``Protective Instrumentation Requirements 
- L. Condenser Vacuum Pump Isolation Function,'' is revised to delete 
line Item No. 1 - ``High Radiation in Main Steam Line Tunnel.''
    Tech. Spec. Table 4.1.1., ``Minimum Check, Calibration and Test 
Frequency For Protective Instrumentation,'' is revised to delete 
Instrumentation Channel No. 13 - ``High Radiation in Main Steam Line.''
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    The objective of the MSLRMs is to provide early indication of 
gross fuel failure. The monitors provide an alarm function, and 
signals that lead to a scram function and [main steam isolation 
valve] MSIV isolation functions. The basis for the MSIV isolation on 
an MSL high radiation signal is to reduce the quantity of fission 
products transported from the reactor vessel to the condenser in the 
event of gross fuel failure. No [design basis accident] DBA takes 
credit for a reactor scram resulting from an MSL high radiation 
signal.
    The proposed change removes all trip functions of the MSLRMs. 
The only modification attendant to this change is the removal of 
contacts derived from the MSLRM logic to the reactor scram, reactor 
isolation and offgas system isolation initiation logic. This change 
does not affect the operation of any equipment having the potential 
to cause a [control rod drop accident] CRDA. Therefore, the 
probability of a CRDA is not increased or in any way affected by the 
proposed change.
    However, the CRDA analysis does take credit for MSIV isolation. 
As discussed above, assuming no MSIV isolation in the event of a 
CRDA, the offsite radiation doses will remain a small fraction of 
the 10 CFR part 100 Reactor Site Criteria.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The function of an MSLRM trip is to detect abnormal fission 
product release and isolate the steam lines, thereby stopping the 
transport of fission products from the reactor to the main 
condenser. No credit is taken for the reactor scram function due to 
the action of these monitors on high radiation in the MSLs in any 
design basis accident. Removing the MSLRMs MSL isolation trip and 
its subsequent reactor scram will not affect the operation of other 
equipment or systems necessary for the prevention or mitigation of 
accidents.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Eliminating the MSLRM trip functions as analyzed in NEDO-31400A 
will result in a potential increase in the margin of safety because 
of:
    a. Improvement in the availability of the main condenser for 
decay heat removal; and,
    b. Elimination of inadvertent reactor scrams and challenges to 
safety systems.
    Therefore, operation of the facility in accordance with the 
proposed changes will not result in a reduction of safety margin.

    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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753.
    Attorney for Licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of Amendment Request: March 9, 1994
    Description of Amendment Request: The proposed amendment would 
modify the Technical Specification to allow a one time exemption from 
certain Appendix J testing. This exemption would extend the interval 
for Type B and C testing until the Unit 2 refueling outage currently 
scheduled for August 1994.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    As stated in 10 CFR 50.92(c), a proposed change does not involve 
a significant hazards consideration if the change does not (1) 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, or (2) the change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, or (3) the change does not involve a 
significant reduction in a margin of safety.

Criterion 1

    The limiting conditions for operation involving containment 
integrity are not altered by this proposed change. The surveillance 
requirement concerning the Type B and C leak rate test is slightly 
relaxed by the proposed change. The function of the components 
affected by this surveillance are to ensure containment integrity. 
Delaying the surveillance approximately two months would not change 
the probability of an accident. Our significant improvement in Type 
B and C leak rate test results, low anticipated leak rate for the 
next surveillance, aggressive corrective actions taken, and 
excellent ILRT [integrated leak rate test] results indicate there is 
no reason to believe that delaying the Type B and C leak rate tests 
approximately two months will cause serious deterioration to these 
components. Furthermore, similar requests by utilities to extend the 
surveillance beyond two years have already been found acceptable by 
the NRC. Therefore, it is concluded that the proposed amendment does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2

    No changes to the limiting conditions for operation for 
containment integrity are proposed as part of this amendment 
request. The proposed change does not involve any physical changes 
to the plant or any changes to plant operations. Thus, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

Criterion 3

    The intent of the Type B and C leak rate surveillance is to 
ensure that containment integrity does not significantly 
deteriorate. This is established by measuring a total leak rate of 
less than 0.60 La. Our significant improvement in Type B and C 
leak rate tests results, aggressive corrective actions taken, and 
excellent ILRT results indicate there is no reason to believe that 
delaying the Type B and C leak rate tests approximately two months 
will cause serious deterioration to these components. The ``As 
Found'' trend of the leak rates over the past three surveillances 
indicate that the leak rate for the next surveillance will be below 
the Appendix J leak rate acceptance criteria. Therefore, it is 
concluded that the proposed amendment does not involve a reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for Licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

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

    Date of Amendment Request: March 23, 1994.
    Description of Amendment Request: The licensee proposed to modify 
Technical Specification Table 3.7-6, Area Temperature Monitoring, by 
creating two zones for the main steam valve building (MSVB) and 
increasing the maximum normal excursion (MNE) temperature limit for 
this area from 120  deg.F to 140  deg.F.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not involve an SHC [significant hazards 
consideration] because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The increase of the MNE temperature from 120  deg.F to 140 
deg.F for the main steam valve building has been evaluated. The 
equipment in the building has been shown to be qualified for 
continuous operation at 140oF. The effect of this temperature change 
has decreased slightly the qualified life of the components in the 
building. For those components with a qualified life of less than 40 
years, they will be replaced as a scheduled maintenance item.
    An engineering review of the MSLB profile for this building was 
conducted and it was concluded that those components required to 
operate post accident, will continue to perform their safety 
function. Therefore, since the equipment will continue to operate as 
designed both during normal conditions and subsequent to a MSLB, the 
probability or consequences of an accident previously evaluated is 
not increased.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The effect of increasing the MNE temperature to 140  deg.F has 
been evaluated and judged acceptable. The possible failure of the 
equipment in this building due to the increase in temperature is no 
more likely than it was before, since the equipment has been shown 
to be qualified to 140  deg.F. Failure of any equipment in this 
building at the new temperature will not create any new accidents or 
consequences that were not considered previously.
    Finally, since there are no changes in the way the plant is 
operated, there is no possibility of an accident of a new or 
different type than previously evaluated due to the proposed change.
    3. Involve a significant reduction in margin of safety.
    The proposed change increases the MNE temperature within the 
MSVB. The equipment in the building has been reviewed to ensure 
operability. There is a slight decrease in the qualified life, but 
this was anticipated and scheduled previously and any such 
replacement of equipment will continue as a maintenance item. A 
review of the MSLB profile was performed for this area and it was 
shown that the required equipment will continue to operate as 
required.

    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.

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

    Date of Amendment Request: March 15, 1994.
    Description of Amendment Request: The proposed amendment would 
include the use of integral fuel burnable absorbers as a method of 
controlling core excess reactivity and maintaining the core power 
distribution within acceptable peaking limitations.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's review is 
presented below:
    1. The proposed amendment would not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    Any fuel containing integral burnable absorbers will be analyzed 
using NRC approved methods and acceptance criteria prior to being 
loaded into Maine Yankee's reactor vessel core. Verification of 
adequate shutdown margin is performed during low power physics testing 
after each refueling. In addition, core physics monitoring is required 
during power operation by Technical Specifications sections 3.10, ``CEA 
Group, Power Distribution, Moderator Temperature Coefficient Limits and 
Coolant Conditions,'' and 3.15 ``Reactivity Anomalies.'' Such testing 
and monitoring ensures adequate margin exists to accommodate the 
anticipated transients and accidents postulated in Maine Yankee's Final 
Safety Analysis Report.
    The licensee therefore concludes that implementation of the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment would not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    A determination of compliance with approved acceptance criteria is 
made for every Maine Yankee fuel reload prior to loading fuel. The use 
of approved methodologies and acceptance criteria ensure that new or 
different accidents will not be created by the use of integral fuel 
burnable absorbers.
    The licensee therefore concludes that implementation of the 
proposed change will not create any or new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment would not involve a significant reduction 
in a margin of safety.
    The safety evaluation performed for each core reload ensures that 
the core design meets appropriate acceptance criteria. Because these 
criteria remain unchanged as approved by the NRC, the margin of safety 
remains the same.
    The licensee therefore concludes that implementation of the 
proposed change would not involve a significant reduction in a margin 
of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room Location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, Maine 04578.
    Attorney for Licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 83 Edison Drive, Augusta, Maine 04336.
    NRC Project Director: Walter R. Butler.

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

    Date of Amendment Request: January 26, 1994.

    Description of Amendment Request: The proposed amendment would 
revise the Technical Specifications and associated Bases to reflect the 
fact that the main steam isolation valves can now be tested at a 
pressure of greater than or equal to Pa (42 psig) thereby 
eliminating the need for the previously granted exemption to certain 
Appendix J testing requirements. The exemption would no longer be 
necessary because of improvements in testing technology.
    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:


    a. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is limited to changes to the surveillance 
testing requirements (test pressure and allowable leakage criteria) 
applicable to the main steam line isolation valves. The proposed 
criteria are equivalent to the current criteria with respect to 
monitoring main steam isolation valve performance to ensure that 
leakage past the valves would be within acceptable limits under 
accident conditions. This surveillance test is performed while the 
plant is in a cold shutdown condition at a time when the main steam 
isolation valves are not required to be operable. Performance of the 
test itself is not an input or consideration in any accident 
previously evaluated, thus the proposed change will not increase the 
probability of any such accident occurring.
    The proposed amendment will not adversely affect the function, 
operation, or reliability of the valves, nor will it diminish the 
capability of the valves to perform as required during an accident. 
There will be no increase in post accident off-site or on-site 
radiation dose, since the adjusted leakage limit is consistent with 
inputs previously established for the dose analyses. The proposed 
amendment is consistent with regulatory requirements (10 CFR Part 
50, Appendix J) and guidance (TER-C5257-30) that has been previously 
reviewed by the NRC and found to be acceptable. Therefore, the 
amendment will not increase the consequences of any accident 
previously evaluated.
    b. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment does not involve any modification to 
plant equipment or operating procedures, nor will it introduce any 
new main steam isolation valve failure modes that have not been 
previously considered. The proposed amendment is limited to a change 
in the surveillance test pressure & acceptance criteria used to leak 
test the valves. This test is performed while the plant is in a cold 
shutdown condition at a time when the valves are not required to be 
operable. We therefore conclude the proposed changes will not create 
the possibility of a new or different kind of accident from any 
accident previously analyzed.
    c. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed amendment will result in the main steam isolation 
valves being subjected to the maximum pressure (Pa, 42 psig) 
calculated to occur under worst case accident conditions, and will 
therefore provide a more realistic and challenging test of valve 
performance under those conditions. The leakage rate criteria for 
the test has been adjusted upward to be commensurate with the higher 
test pressure, but this does not represent any increase in actual 
leakage under accident conditions. On-site and off-site dose 
analyses will not be affected. The proposed amendment does not 
involve any change in operability requirements or limiting 
conditions for operation beyond the replacement of the old test 
pressure & acceptance criteria with equivalent criteria consistent 
with 10 CFR Part 50, Appendix J, NUREG-1433, and TER-C5257-30. Based 
on these considerations, we conclude the proposed amendment will not 
involve a significant reduction in the margin of safety.


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for Licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

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

    Date of Application for Amendment: April 6, 1994.
    Description of Amendment Request: The amendment would reflect the 
incorporation of the end-of-cycle Minimum Critical Power Ratio 
Recirculation Pump Trip (MCPR-RPT) system and the replacement of the 
Reactor Recirculation System (RRS) Motor Generator (M-G) Sets with 
solid state adjustable speed drives (ASDs).
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


    (1) The proposed change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The addition of the end-of-cycle MCPR-RPT System, which utilizes 
ASDs, will not have a significant increase in the probability or 
consequences of an accident previously evaluated.
    The end-of-cycle MCPR-RPT System has been designed to 
appropriate standards and specifications to ensure that the ability 
of the plant to mitigate the effects of accidents is maintained. 
Additionally, the MCPR-RPT System has been analyzed such that no new 
accident initiators will be created such that the probability of an 
accident previously evaluated will not increase.
    No new challenges to the reactor coolant pressure boundary will 
result from the incorporation of the end-of-cycle MCPR-RPT System 
which could result in an increase in the consequences of an 
accident. All engineered safety features will function as described 
in the PBAPS UFSAR [Peach Bottom Atomic Power Station Updated Final 
Safety Analysis Report] in order to mitigate the consequences of 
accidents previously evaluated in the PBAPS UFSAR. Additionally, all 
fission product barriers and safety margins will be maintained.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The end-of-cycle MCPR-RPT System, which utilizes ASDs, has been 
designed to appropriate standards and specification to ensure that 
no new sequence of events or failure modes will occur such that a 
transient event will escalate into a new or different type of 
accident.
    The software used in the digital system of the ASDs is not 
subject to the verification and validation requirements discussed in 
the NRC memorandum dated July 1, 1991, from A. C. Thadoni [sic] 
[Thadani] (NRC) to S. A. Varga (NRC) and B. A. Bolger [sic] [Boger] 
(NRC), because this equipment is neither safety-related nor 
important to safety. There is no software used in the trip circuit 
of the end-of-cycle MCPR-RPT System, except for the ASDs. 
Additionally, the design of the modification will assure that the 
new equipment EM emissions will not cause inadvertent operation of 
existing plant equipment and that harmonic filters have been 
incorporated to minimize electrical noise on the 13kV input power 
buses.
    (3) The proposed change does not result in a significant 
reduction in the margin of safety.
    The incorporation of the end-of-cycle MCPR-RPT System, which 
utilizes ASDs, will not result in a reduction in the margin of 
safety. All safety margins will be maintained.
    The end-of-cycle MCPR-RPT System will aid in protecting the 
integrity of the fuel barrier by tripping the recirculation pumps 
early in the pressurization phase of the load rejection with no 
bypass event, the turbine trip with no bypass event, and the 
feedwater controller failure--maximum demand event. The early 
tripping of the recirculation pumps will introduce negative void 
reactivity thus reducing reactor power and maintaining safety 
margins. The end-of-cycle MCPR-RPT System will ensure CPR safety 
margins which protect fuel barrier integrity.
    General Electric has performed a qualitative assessment of 
transients that would be impacted as a result of replacing the M-G 
Sets with ASDs. General Electric concluded that the faster coastdown 
of the recirculation pumps during a Loss of Coolant Accident (LOCA) 
due to the removal of the M-G Set inertia may slightly increase the 
peak clad temperature during this event. This increase is expected 
to be less than 50 deg.F. The small increase will not exceed the 
2200 deg.F peak cladding temperature regulatory limit. No design or 
safety limit will be exceeded.
    The replacement of the M-G Sets with the ASDs will not impact 
the recirculation flow controller failure--increase flow transient. 
The UFSAR analysis assumes a 25%/sec rate of increase. The ASD 
control system will include rate limiters that prevent a pump speed 
increase greater than 25%/sec in the event of a failure. Thus, the 
consequences of this transient remain bounded and safety margins 
will be maintained.
    The ASDs will also allow a ``soft start'' of the recirculation 
pumps with the recirculation discharge valves closed prior to pump 
start and a gradual increase in pump speed. This results in a 
gradual change in core flow. Thus, the response to a startup of an 
idle recirculation pump remains bounded by the transient analysis 
and safety margins will be maintained in the transient analyses.
    Changes to the fire protection equipment will still maintain the 
capability to shutdown the plant in the event of a fire.


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, 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: March 28, 1994.
    Description of Amendment Request: The proposed Technical 
Specifications (TS) changes relocate the TS fire protection 
requirements to the Updated 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 changes are administrative in nature and are 
consistent with the guidance provided in NRC GL's [Generic Letters] 
86-10 and 88-12. 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 
consistent with the guidance provided in NRC GL's 86-10 and 88-12. 
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.
    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.

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

    Date of Amendment Request: January 21, 1994.
    Description of Amendment Request: These amendments would revise 
Technical Specifications 3.8.2.3 for both Salem Unit 1 and Salem Unit 2 
to include the battery acceptance criteria, corresponding allowed 
outage times and additional surveillance requirements recommended in 
NUREG-1431, Standard Technical Specifications--Westinghouse Plants.
    TS 3.8.2.4 ``125 Volt D.C. Distribution--Shutdown'' would also be 
indirectly affected by these changes because it refers to the 
surveillance requirements of TS 4.8.2.3.2 to demonstrate the battery 
and chargers Operable.
    In addition, Salem Unit 1 TS 3.8.2.3 Limiting Condition for 
Operation (LCO) would be revised to define the specific battery charger 
required for each train. Salem Unit 1 TS 3.8.2.3 Action Statement would 
also be revised to restrict the use of the backup battery charger to a 
period not to exceed 7 days.
    Additionally, the Unit 1 action statement for an inoperable 125 
volt DC bus would be modified to add the requirement that the bus also 
be energized.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not alter plant configuration or 
operation. The proposed changes do not invalidate any of the 
parameters assumed in the plants UFSAR Design Basis Accident or 
Transient Analyses. The proposed changes provide additional guidance 
to be used to ensure operability of the safety related batteries. 
New surveillance requirements and specific battery cell parameters 
offer improved monitoring of the battery status. The new guidance 
and surveillance requirements are consistent with the 
recommendations of NUREG-1431, Standard Technical Specifications--
Westinghouse Plants, and current industry recommendations.
    The changes to the Unit 1 LCO and corresponding Action Statement 
restrict the use of the backup battery charger, thereby limiting the 
amount of time that one AC Vital bus is allowed to power the 
chargers of more than one DC train. This change brings the TS for 
both Units into agreement and results in a more conservative Unit 1 
TS.
    Therefore, the probability or consequences of an accident 
previously evaluated are not increased by the proposed change.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not introduce any design or physical 
configuration changes to the facility or change the method by which 
any safety-related system performs its function. The proposed 
changes are consistent with the recommendations of NUREG-1431, 
Standard Technical Specifications--Westinghouse Plants. Therefore, 
the proposed changes will not increase the possibility of a new or 
different kind of accident from any accident previously identified.
    3. Does not involve a significant reduction in a margin of 
safety.
    The proposed changes do not alter the manner in which safety 
limits or limiting safety system setpoints are determined. The new 
cell parameter table and additional surveillance requirements 
provide improved means to monitor and evaluate overall battery 
performance. Therefore, the proposed changes do not involve a 
significant reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for Licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Charles L. Miller.

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

    Date of Amendment Request: 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 (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 
increasing the maximum tested charging/safety injection pump flow rate 
from 680 gallons per minute to 700 gallons per minute.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of 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, 
including the CHG/SI [charging and safety injection] system with 
maximum pump flows up to 700 gpm, 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 [non-loss-of-coolant accident]
--Large Break and Small 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, revised operating conditions, and higher 
allowable CHG/SI pump flows 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, revised operating conditions, and 
higher allowable CHG/SI pump flows 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. 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 10 CFR 50.46. The calculated 
doses resulting from a SGTR event will continue to remain within a 
small fraction of the 10 CFR 100 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 of 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: William H. Bateman.

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

    Date of Amendment Request: March 12, 1993.
    Description of Amendment Request: The proposed amendment would 
revise Technical Specification Table 3.3.7.1-1, to clarify the actions 
to be taken if the control room ventilation radiation monitor is not 
operable.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change clarifies Technical Specification 3.3.7.1, 
``Radiation Monitoring Instrumentation'' by revising Action 72 (for an 
inoperable Control Room Ventilation Radiation Monitor) to remove 
several inconsistencies between it and Action 3.7.2.b.2 of the Control 
Room Emergency Recirculation System Specification. Revised Action 72 
simply makes the two Specifications more consistent by incorporating 
alternative compensatory measures that the operators may take after the 
Control Room Ventilation Radiation Monitor has been inoperable for more 
than seven days. The proposed Action would retain the choice of 
initiating at least one train of the Control Room Emergency 
Recirculation System, while providing a second option to take which 
would depend on the current Operational Condition. In Operational 
Conditions 4, 5 and * * * the current Specification 3.3.7.1 Action 72 
does not contain the provisions of the Control Room Emergency 
Recirculation System Action 3.7.2.b.2 which directs the Operators to 
suspend performance of Core Alterations, handling of irradiated fuel 
and operations with a potential for draining the reactor vessel instead 
of initiating the Control Room Emergency Recirculation System. This 
inconsistency between the two specifications has caused compliance 
difficulties; therefore, the proposed Action adds this alternative. 
Also, in Operational Conditions 1, 2 and 3 a shutdown provision is 
being added. The other changes are editorial, in order to clarify the 
applicability of the proposed alternative compensatory measures, to be 
consistent with PNPP-specific terminology, and to be more consistent 
with Action b of Specification 3.7.2.
    In summary, there is no change in the probability or consequences 
of any accident since the revision of Specification 3.3.7.1 Action 72 
is simply proposed in order to achieve consistency with the current 
Action 3.7.2.b.2. Incorporation of the already approved 3.7.2.b.2 
compensatory measures to suspend possible radiation accident initiating 
activities provides an alternative which would actually reduce the 
probability of occurrence of a previously analyzed accident, and would 
have no adverse effect on accident consequences. None of the proposed 
changes to the clarified action, including the editorial changes, 
involves a change to the design of the plant, nor the operational 
characteristics of any plant system, nor the procedures by which the 
Operators run the plant.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No design changes are being made that would create a new type of 
accident or malfunction, and the methods and manner of plant operation 
remains unchanged. The proposed revisions to Action 72 will remove 
several inconsistencies between the two Specifications by providing 
consistent actions within the Radiation Monitoring Instrumentation 
Specification with those currently existing in the Control Room 
Emergency Recirculation System Specification and provide an additional 
shutdown requirement in Operational Conditions 1, 2 and 3. The other 
changes to Action 72 are editorial, and therefore cannot affect 
accident initiation parameters. The instrument to which Action 72 
applies (the Control Room Ventilation Radiation Monitor (Noble Gas)) 
simply serves as a supporting instrumentation channel for the Control 
Room Emergency Recirculation System, therefore no new or different kind 
of accident can be created.
    3. The proposed changes do not involve a significant reduction in a 
margin of safety.
    The proposed change to Specification 3.3.7.1 Action 72 simply makes 
the two Specifications more consistent by making the Action for a 
supporting instrumentation channel, the Control Room Ventilation 
Radiation Monitor (Noble Gas), more consistent with those of the 
supported system Specification, the Control Room Emergency 
Recirculation System. A shutdown requirement is also being added if the 
operators should choose not to initiate the supported system in 
Operational Conditions 1, 2, and 3. Since the Actions of the two 
Specifications will now correspond, the margin of safety as currently 
exists today for the governing Specification (the Control Room 
Emergency Recirculation System Specification) is maintained and the 
proposed changes do not therefore reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for Licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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 Amendment Request: March 18, 1994.
    Description of Amendment Request: The proposed amendment would 
revise TS 2.1.2 (Reactor Core), TS 2.2.1 (Reactor Protection System 
Setpoints), Bases 2.1.1 and 2.1.2 (Reactor Core), Bases 2.2.1 (Reactor 
Protection System Instrumentation Setpoints), TS 3.2.2 and 3.2.3 (Power 
Distribution Limits), Bases 3/4.2 (Power Distribution Limits), and TS 
6.9.1.7 (Administrative Controls, Core Operating Limits Report). This 
amendment would remove cycle-specific limits from TS and relocate them 
in the Core Operating Limits 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, indicating that the proposed 
changes would:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
assumptions or probabilities are affected by the proposed relocation 
of cycle-specific core operating limits to the Core Operating Limits 
Report.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated. The proposed changes do not affect 
any equipment, accident conditions, or assumptions which could lead 
to a significant increase in radiological consequences.
    2a. Not create the possibility of a new kind of accident from 
any accident previously evaluated because no new accident initiators 
are introduced by these proposed changes.
    2b. Not create the possibility of a different kind of accident 
from any accident previously evaluated because no different accident 
initiators are introduced by these proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes only relocate cycle-specific core 
operating limits to the Core Operating Limits Report; they do not 
allow less conservative operating limits. The analytical methods to 
be used in the determination of cycle-specific core operating limits 
are previously approved by the NRC. The same margin of safety 
provided in the current Technical Specifications will continue to be 
maintained.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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 Amendment Request: March 30, 1994.
    Description of Amendment Request: The proposed amendment would add 
a new TS Limiting Condition for Operation 3/4.4.12, Pilot Operated 
Relief Valve and Block Valve, and would include associated Bases and 
Surveillance Requirements.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, indicating that the proposed 
additions and changes would:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. Automatic actuation of the PORV is not assumed 
to mitigate the consequences of a design basis accident as described 
in Chapter 15 of the USAR. The proposed changes will continue to 
ensure the PORV and block valves are available to perform their 
functions when required to do so. Therefore, it can be concluded 
that the proposed changes do not involve a significant increase in 
the probability of an accident previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate accident conditions or assumptions used in evaluating the 
radiological consequences of an accident.
    2a. Not create the possibility of a new kind of accident from 
any accident previously evaluated because the proposed changes do 
not delete any function previously provided by the PORV nor has the 
possibility of inadvertent opening been increased. No new types of 
failures or accident initiators are introduced by the proposed 
changes.
    2b. Not create the possibility of a different kind of accident 
from any accident previously evaluated because no new failure modes 
have been defined for any plant system or component important to 
safety, nor has any new limiting single failure been identified as a 
result of the proposed changes. No different accident initiators or 
failure mechanisms are introduced by the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes continue to ensure the availability of 
the PORV and block valve when called upon to perform their function 
and will not impact any safety analysis assumptions.

    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 Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of Amendment Request: February 14, 1994.
    Brief Description of Amendments: The proposed amendment would 
revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 
technical specifications to increase the Unit 2 boron concentration for 
the refueling water storage tank (RWST) and the emergency core cooling 
system (ECCS) accumulators to support Unit 2 operation with extended 
fuel cycles. These changes are applicable to Unit 2 only and are 
identical to those previously approved for Unit 1.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed change would not increase the probability or 
consequences of a previously evaluated accident.
    The proposed changes are related to the boron concentration in the 
RWST and ECCS accumulators. This increased concentration does not 
constitute a change expected to increase the probability of a 
previously evaluated accident. The means by which the proposed changes 
might result in increased radiological consequences of various 
accidents are discussed below.
    The higher boron concentration may result in increased probability 
of equipment failure following an accident due to in-containment or in-
process equipment being exposed to a more severe post-accident 
environment. The general chemical properties of the slightly higher 
boron concentration fluid indicates no mechanism that would result in 
an appreciable increase in the component failure rate. While the 
corrosive nature of the fluid will increase, this increase will be only 
minimal. Thus, there is no significant increase in the consequences of 
any accident due to an increase in the probability of equipment 
failure.
    The changes in containment spray and sump solution pH may change 
the radioisotope removal and partition characteristics. While some 
relevant characteristics are affected, the resulting limiting 
coefficient values associated with the pH changes are bounded by the 
values used in the design calculations for CPSES. Thus, no adverse 
impact of the radiological consequences arising from this mechanism has 
been identified.
    The impact of the containment spray, with a lower pH, upon the 
combustible gas production rate was also evaluated. No mechanism for 
increased combustible gas production was identified.
    The higher boron concentration could have an adverse impact on the 
inadvertent actuation of the ECCS event. Although the timing of the 
sequence of events may be affected, the departure from nucleate boiling 
ratio continues to increase from its initial value throughout the 
event. On the basis of its review of this event, the licensee has 
identified no changes in the event probability or consequences; 
however, the continued validity of this conclusion will be reconfirmed 
by the licensee on a cycle-specific basis.
    2. The proposed change would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed change only changes the allowable boron concentration. 
No new or different accident sequences have been identified. 
Furthermore, the licensee has reviewed the heat tracing requirements 
and determined that there are no additional requirements resulting from 
the boron concentration increase. There are no previously unconsidered 
failure mechanisms.
    3. The proposed change would not involve a significant reduction in 
the margin of safety.
    The decrease in the containment spray and sump solution pH could be 
expected to result in higher airborne iodine concentrations. The 
accident source terms could be impacted by variations in the iodine 
spray removal and partition factors. A comparison of the coefficients 
for the minimum equilibrium containment sump solution pH to those used 
in the CPSES design analyses indicated that the expected coefficient 
values would remain bounded by the values used in the previous 
analyses. Thus, no significant reduction in the margin of safety has 
been identified.
    Based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves no significant hazards 
consideration.
    Local Public Document Room Location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.
    Attorney for Licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, NW., suite 1000, Washington, DC 20036.
    NRC Project Director: Suzanne C. Black.

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

    Date of Amendment Request: February 14, 1994.
    Brief Description of Amendments: The proposed amendment will revise 
the Comanche Peak Steam Electric Station, Units 1 and 2, technical 
specifications to be consistent with the new 10 CFR part 20.
    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 changes do not involve a significant increase in 
the probability or consequences of a previously evaluated accident.
    The proposed revisions to the liquid and gaseous effluent 
release limits will not change the type or amount of effluent 
released nor will there be an increase in individual or cumulative 
dose. The changes will result in levels of radioactive materials in 
effluents being maintained ALARA [as low as reasonably achievable] 
and comply with 10 CFR 50.36a and 10 CFR 50 Appendix I. The change 
to the high radiation area dose measurement distance will ensure 
that high radiation areas are conservatively posted per 10 CFR 
20.1601(a)(1) and provide controls to minimize individual dose. The 
changes do not impact the operation or design of any plant 
structure, system or component. Other proposed changes are 
administrative only. Therefore, the proposed changes do not involve 
an increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not affect the plant design or operation 
nor do they result in a change to the configuration of any 
equipment. No change is proposed that will change the type or 
quantity of effluents released off site or change the source terms 
available for release. Therefore, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not change the type or increase the 
amount of effluents released offsite. No change in the methodology 
used to control radioactive waste or radiological environmental 
monitoring is proposed. Control of radioactive effluents and 
effluent monitor setpoints will be based on current dose to the 
public limitations. Under the proposed change, high radiation area 
measurements are more conservative and will not result in an 
increase in individual or cumulative occupational radiation 
exposures. Compliance with the limits of the revised 10 CFR 20.1301 
will be demonstrated by operating within the limits of 10 CFR 50, 
Appendix I and 40 CFR 190. Therefore, these changes do not reduce 
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: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.
    Attorney for Licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street NW., suite 1000, Washington, DC 20036.
    NRC Project Director: Suzanne C. Black.

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

    Date of Amendment Request: February 14, 1994.
    Brief Description of Amendments: The proposed amendment would 
revise the Comanche Peak Steam Electric Station Units 1 and 2 technical 
specifications by reducing the frequency of reports for radiological 
effluents from semiannual to annual, and change the due date from 
within 60 days after January 1 and July 1 to prior to May 1.
    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 changes do not involve a significant increase in 
the probability or consequences of a previously evaluated accident.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The amendment involves only changes of reporting 
frequency and due date requirements for radiological effluent 
release reporting. These changes are administrative in nature and do 
not affect safe operation of the plant; therefore, accident 
probabilities or consequences are unaffected.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment is administrative in nature and 
does not involve any changes to plant design of configuration. For 
this reason, it will not create the possibility of a new or 
different kind of accident.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed amendment does not involve a significant reduction 
in the margin of safety. The proposed amendment only changes the 
reporting frequency and due date requirements for radiological 
effluent release reporting. The reporting requirements for 
radiological effluent releases are administrative changes: 
therefore, there is not 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: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.
    Attorney for Licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street NW., suite 1000, Washington, DC 20036.
    NRC Project Director: Suzanne C. Black.

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

    Date of Amendment Request: March 30, 1994.
    Description of Amendment Request: The proposed changes would revise 
the North Anna Power Station, Units No. 1 and No. 2 (NA-1&2) Technical 
Specifications (TS). Specifically, the proposed changes would revise 
the High Head Safety Injection (HHSI) flow balance surveillance 
requirements by removing specific numerical values. The numerical 
values would be replaced with broader requirements to ensure that the 
HHSI flow rates meet the loss of coolant accident (LOCA) analysis 
acceptance criteria and pump runout limits. The NA-1&2 TS 4.5.2.h 
requires a surveillance test of the HHSI system following the 
completion of any modification to the Emergency Core Cooling System 
(ECCS) subsystems that could alter the subsystem flow characteristics. 
The current surveillance criteria specify values for the sum of the 
injection line flow rates, excluding the highest flow rate, and the 
total pump flow rate. These correspond to requirements for the safety 
analysis flow input and the HHSI pump runout limit, respectively.
    The HHSI test acceptance criteria in the current TS are very narrow 
because of the various system physical and technical constraints that 
need to be considered in the flow balance testing. These acceptance 
criteria may also be more restrictive than required by either the LOCA 
analysis or the actual pump runout requirements. For example, the LOCA 
analysis contains input conservatisms that could be used to offset a 
reduction in the required HHSI flow while still meeting the 10 CFR 
50.46 LOCA acceptance criteria. The proposed TS changes would permit 
the use of additional available margin, while maintaining a strong 
technical linkage between the measured system performance and the 
safety analysis. Although these proposed TS changes remove the 
numerical values from TS 4.5.2.h, neither the methodology nor the 
acceptance criteria for LOCA analysis are affected.
    Basis for Proposed No Significant Hazards Consideration 
Determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specification changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
Technical Specification changes continue to require that with one 
HHSI pump running, the sum of the flows through the two lowest 
branch lines shall be [greater than or equal to] the minimum HHSI 
flow required by the safety analysis and that the total HHSI pump 
flow rate shall be [less than or equal to] the evaluated HHSI pump 
runout limit.
    Likewise, the consequences of the accidents previously evaluated 
will not increase as a result of the proposed Technical 
Specification changes. The system performance will remain bounded by 
the safety analysis for all postulated conditions. The safety 
analysis will continue to be performed and evaluated in accordance 
with the requirements of 10 CFR 50.59 and 10 CFR 50.46.
    2. Create the possibility of a new or different kind of accident 
or malfunction from any previously evaluated. The proposed Technical 
Specification changes will not affect the capability of the HHSI 
System to perform its intended function. The proposed Technical 
Specification changes are bounded by the existing safety analysis 
and do not involve operation of plant equipment in a different 
manner from which it was designed to operate.
    Since a new failure mode is not created, a new or different type 
of accident or malfunction is not created.
    3. Involve a reduction in a margin of safety. The system 
performance will continue to bound the flow rates specified in the 
safety analysis, therefore safety margins are not reduced.

    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.

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

    Date of Application for Amendments: March 21, 1994.
    Description of Amendment Requests: The proposed amendments would 
permit continued activities at all four units with main steam Code 
safety valve tolerances of plus or minus 3% until the lift setpoints 
can be reset to within plus or minus 1%.
    Date of Publication of Individual Notice in Federal Register: March 
29, 1994 (59 FR 14685).
    Expiration Date of Individual Notice: April 29, 1994.
    Local Public Document Room Location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

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

    Date of Amendment Request: March 24, 1994.
    Description of Amendment Request: The proposed amendment would 
revise section 6.0 (Administrative Controls). Specifically, the plant 
staff requirement (specified in Technical Specification (TS) 6.2.2.i) 
would be revised to temporarily allow the operations manager to have 
held a senior reactor operator (SRO) license at a pressurized water 
reactor (PWR) other than Indian Point 3. The TS currently requires the 
operations manager to have or have held an SRO license at Indian Point 
3 only. This proposed change is needed to allow management changes at 
the facility in an effort to improve overall performance. The proposed 
changes would be in effect for a period ending 3 years after restart 
from the 1993/1994 Performance Improvement Outage.
    Date of Publication of Individual Notice in Federal Register: April 
1, 1994 (59 FR 15464).
    Expiration Date of Individual Notice: May 3, 1994.
    Local Public Document Room Location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for Licensee: Charles M. Pratt, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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.

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

    Date of Application for Amendments: December 2, 1993.
    Brief Description of Amendments: The amendments will modify TS 3/
4.6.1.2 by removing the schedular requirements for a Type A (overall 
integrated containment leakage rate) test to be performed specifically 
at 40 plus or minus 10-month intervals and replacing these requirements 
with a requirement to perform Type A testing in accordance with 
Appendix J to 10 CFR part 50.
    Date of Issuance: April 6, 1994.
    Effective date: April 6, 1994.
    Amendment Nos.: 73, 59, and 45.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of Initial Notice in Federal Register: January 5, 1994 (59 FR 
616) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 6, 1994.
    No significant Hazards Consideration Comments Received: No.
    Local Public Document Room Location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of Application for Amendments: August 27, 1993, as 
supplemented March 11, 1994.
    Brief Description of Amendments: The amendments revise the Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications 
(TSs) by removing the list of containment isolation valves in Table 
3.6-1. The amendments also make accompanying changes to various TSs and 
to the TS Bases. These amendments are a ``line-item'' TS improvement 
and follow the guidance of Generic Letter 91-08, ``Removal of Component 
Lists From Technical Specifications.''
    Date of issuance: April 7, 1994.
    Effective Date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 187 and 164
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: September 29, 1993 (58 
FR 50966) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of Application for Amendment: June 7, 1993, August 9, and 
December 10, 1993.
    Brief Description of Amendment: This amendment revises the 
Technical Specification (TS) to support a 24-month fuel cycle. The TS 
changes include extending surveillance intervals and adjusting 
setpoints as justified in the Safety Evaluation.
    Date of Issuance: April 6, 1994.
    Effective Date: April 6, 1994.
    Amendment No.: 151.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: March 16, 1994 (59 FR 
2863) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of Application for Amendments: March 26, 1993.
    Brief Description of Amendments: The amendments modify the trip 
level settings for the Isolation Condenser and High Pressure Core 
Injection System Steam lines to more conservative values. In addition, 
the proposed amendments revise the ECCS Low-Low Water Level initiation 
trip setting to a more conservative number.
    Date of Issuance: April 5, 1994.
    Effective Date: April 5, 1994.
    Amendment Nos.: 126 and 120.
    Facility Operating License Nos. DPR-19 and DPR-25. The amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10002) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 5, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Morris Public Library, 604 
Liberty Street, Morris, Illinois 60450.

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

    Date of Application for Amendments: January 24, 1994.
    Brief Description of Amendments: The amendments implement line item 
5.9 of Generic Letter 93-05, ``Line-Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation'', which provided recommendations for deleting the 
requirement to perform response time testing where the required time 
corresponds to the diesel start time.
    Date of issuance: April 7, 1994.
    Effective Date: April 7, 1994.
    Amendment Nos.: 98 and 82.
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: February 16, 1994 (59 
FR 7686). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

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

    Date of Application for Amendments: February 22, 1993 as 
supplemented August 16, 1993.
    Brief Description of Amendments: The amendments allow continued 
operation of one unit for a period of seven days while the common plant 
(Division 1) emergency diesel generator (``O'' DG) is out of service 
for the performance of specified Technical Specification surveillance 
requirements and the performance of planned maintenance and/or 
modification work. Also, the amendments clarify Surveillance 
Requirement 4.8.1.1.2.a.7 to allow an emergency diesel generator to 
remain Operable with only one air start subsystem pressurized.
    Date of Issuance: April 11, 1994.
    Effective Date: April 11, 1994.
    Amendment Nos.: 99 and 83.
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: July 7, 1993 (58 FR 
36430) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 11, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

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

    Date of Application for Amendments: September 28, 1993, as 
supplemented February 17, 1994.
    Brief Description of Amendments: The amendments delete the portion 
of the 18-month surveillance requirement contained in Technical 
Specification (TS) 4.5.2.d associated with verifying that the decay 
heat removal system suction isolation valves automatically close on a 
reactor coolant system pressure signal. Also, an obsolete footnote to 
TS 4.5.2.e is being deleted. This footnote is no longer necessary since 
the first Unit 1 refueling outage is complete.
    Date of Issuance: April 4, 1994.
    Effective Date: April 4, 1994.
    Amendment Nos.: 117 and 111.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10004) The February 17, 1994, letter provided clarifying information 
that did not change the scope of the initial September 28, 1993, 
application and initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 4, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of Application for Amendment: July 22, 1993, as supplemented 
by letter dated October 20, 1993.
    Brief Description of Amendment: The amendment removed the cycle-
specific variables from the Technical Specifications (TSs) and 
controlled them under a new document called the Core Operating Limits 
Report (COLR), in accordance with Generic Letter 88-16.
    Date of Issuance: April 11, 1994.
    Effective Date: April 11, 1994.
    Amendment No.: 157.
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: September 1, 1993 (58 
FR 46230). The additional information contained in the supplemental 
letter dated October 20, 1993, was clarifying in nature and, thus, 
within the scope of the initial notice and did not affect the staff's 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 11, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of Application for Amendment: January 13, 1994.
    Brief Description of Amendment: The amendment requested the removal 
of the temporary technical specification limit on the number of spent 
fuel assemblies that may be stored in the spent fuel pool at Grand Gulf 
Nuclear Station pending licensee verification of the adequacy of the 
spent fuel pool heat removal capability.
    Date of Issuance: April 4, 1994.
    Effective Date: April 4, 1994.
    Amendment No: 113.
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10006) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 4, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, Post Office Box 1406, S. Commerce at Washington, Natchez, 
Mississippi 39120.

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: April 20, 1993.
    Brief Description of Amendments: These amendments delete the lead/
lag compensator term on the measured reactor coolant system loop 
temperature difference from the overtemperature and overpower Delta T 
reactor trip functions.
    Date of Issuance: April 4, 1994.
    Effective Date: April 4, 1994.
    Amendment Nos. 161 and 155.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: June 9, 1993 (58 FR 
32383) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 4, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of Application for Amendments: September 20, 1993.
    Brief Description of Amendments: The amendments revise the Units 1 
and 2 Channel Functional Test frequency from quarterly to once per 18 
months for the scram discharge volume float type level switches.
    Date of Issuance: April 15, 1994.
    Effective Date: To be implemented within 60 days from the date of 
issuance.
    Amendment Nos.: 193 and 133.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: October 27, 1993 (58 FR 
57852) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 15, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

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.
    Brief Description of Amendment: The amendment added new Technical 
Specifications, 3/4.10.6 and 3/4.10.7, to the Special Test Exceptions 
section. TS 3/4.10.6 allows the restart of Unit 2 with expired 
calibrations on the core exit thermocouples (CET) and the reactor 
coolant system (RCS) resistance temperature detectors (RTD) by setting 
aside the affected limiting conditions for operation (LCOs) until the 
calibrations are complete. This is a one-time only change that is valid 
during the third refueling outage for Unit 2 until the calibrations are 
complete. TS 3/4.10.7 adds a new technical specification to allow the 
ascension to 75 percent rated thermal power with an expired precision 
heat balance reactor coolant flow measurement. This change is effective 
only for Unit 2, Cycle 4, until the surveillance requirement is 
completed.
    Date of Issuance: April 1, 1994.
    Effective Date: April 1, 1994, to be implemented within 10 days of 
issuance.
    Amendment No.: Amendment No. 48.
    Facility Operating License No. NPF-80. Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: February 16, 1994 (59 
FR 7690) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 1, 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.

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

    Date of Application for Amendment: January 21, 1994.
    Brief Description of Amendment: The amendment revises Technical 
Specification 4.6.3 (Emergency Power Sources), to eliminate unnecessary 
testing of an operable emergency diesel generator (EDG) when the 
redundant EDG becomes inoperable. This amendment is intended to 
increase EDG reliability and the overall level of plant safety by 
reducing the stresses on the EDGs caused by unnecessary testing. This 
amendment also eliminates the requirement to load the operable EDG with 
the offsite network when it is tested with one EDG inoperable.
    Date of Issuance: April 6, 1994.
    Effective Date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 147.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10009) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of Application for Amendment: November 18, 1993.
    Brief Description of Amendment: The amendment revises the setpoints 
for the degraded voltage relays for the 4.16kV Power Boards 102 and 103 
as specified in Technical Specification Table 3.6.2i. The setpoints 
have been revised from 3580 volts  3 seconds to 3705 volts 
> 3.4 seconds and < 60 seconds. This change has been made in response 
to findings of the NRC's Electrical Distribution System Functional 
Inspection conducted at Nine Mile Point Nuclear Station Unit No. 1 from 
September 23, 1991, to October 25, 1991.
    Date of Issuance: April 7, 1994.
    Effective Date: As of the date of issuance to be implemented prior 
to startup from the next refueling outage.
    Amendment No.: 148.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of Initial Notice in Federal Register: December 22, 1993 (58 
FR 67851).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendment request: October 28, 1993.
    Description of Amendment Request: The amendment implements 13 of 47 
line item Technical Specification (TS) improvements recommended by 
Generic Letter 93-05. Most of the changes revise the allowable time 
intervals for performing certain Surveillance Requirements (SR) on 
various plant components during power operation or delete the 
requirement entirely or under certain conditions. One change modifies 
testing requirements identified in an ACTION statement. Specifically, 
the amendment modifies Surveillance Requirements 4.1.3.1.2, 4.6.4.1, 
4.3.2.1 (Table 4.3-2, Functional Unit 3.c.4), 4.3.3.1 (Table 4.3-3, 
Functional Units 1 through 6), 4.4.6.2.2, 4.4.11.1, 4.4.3.2, 4.5.1.1.1, 
4.5.1.1.2, 4.5.2, 4.6.2.1, 4.6.4.2, 4.7.1.2.1, and the ACTION 
statements in Technical Specification 3.8.1.1.
    Date of Issuance: April 7, 1994.
    Effective Date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 30.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: February 2, 1994 (59 FR 
4942). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 7, 1994.
    No Significant Hazards Consideration Comments Received: No.
    Local Public Document Room Location: Exeter Public Library, 47 
Front Street, Exeter, New Hampshire 03833.

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

    Date of Application for Amendment: July 7, 1993.
    Brief Description of Amendment: The amendment changes Technical 
Specification 3.6.D, ``Primary System Boundary, Coolant Leakage,'' and 
the corresponding surveillance requirements. The amendment adds a 
clause to make the operability requirements of leakage measurement 
instruments applicable only when irradiated fuel is in the reactor and 
reactor water temperature is above 212 deg.F. With regards to leakage 
measurement instruments, it is now required that leak rate measurements 
be made once per 12 hours. In addition, instruments must be restored to 
operable status within 30 days or else shutdown would be required. 
Operability requirements for the drywell particulate radioactivity 
monitoring system are now addressed. Surveillance requirements 
regarding primary containment atmosphere, identified and unidentified 
leakage of reactor coolant, and performance of a sensor check for the 
primary containment sump leakage measurement system are changed to once 
per shift, not to exceed 12 hours.
    Date of Issuance: April 15, 1994.
    Effective Date: April 15, 1994.
    Amendment No.: 87.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: August 4, 1993 (58 FR 
41507) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 15, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of Application for Amendment: November 24, 1993, and 
supplemented by letters dated January 7, and February 14, 1994.
    Brief Description of Amendment: The amendment raised the authorized 
power level from the 3293 MWt to a new limit of 3441 MWt. The amendment 
also changed the Technical Specifications to implement uprated power 
operation.
    Date of Issuance: April 11, 1994.
    Effective Date: As of its date of issuance and is to be implemented 
prior to startup in Cycle 7, currently scheduled to occur May 21, 1994.
    Amendment No.: 103.
    Facility Operating License No. NPF-22. This amendment revised the 
Technical Specifications and the License.
    Date of Initial Notice in Federal Register: December 22, 1993 (58 
FR 67852)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 11, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

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

    Date of Application for Amendments: April 16, 1993.
    Brief Description of Amendments: The amendments revised the 
Technical Specifications to conform to the NRC staff positions on 
Inservice Inspection and on monitoring of unidentified leakage in 
Generic Letter 88-01, ``NRC Position On Intergranular Stress Corrosion 
Cracking In BWR Austenitic Stainless Steel Piping''.
    Date of Issuance: April 15, 1994.
    Effective Date: April 15, 1994.
    Amendment Nos.: 134 and 104.
    Facility Operating License Nos. NPF-14 and NPF-22. These amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: May 12, 1993 (58 FR 
28058). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 15, 1994.
    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, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric Company, 
Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit 
Nos. 2 and 3, York County, Pennsylvania

    Date of Application for Amendments: May 21, 1992
    Brief Description of Amendments: The amendments revised the 
expiration dates from January 31, 2008, for Units 2 and 3, to August 8, 
2013, for Unit 2, and July 2, 2014, for Unit 3. The original expiration 
date is 40 years from the date of issuance of the Construction Permit 
for both units. The revised dates are 40 years from the date of 
issuance of the respective Operating Licenses (i.e., August 8, 1973 for 
Unit 2 and July 2, 1974 for Unit 3).
    Date of Issuance: March 28, 1994.
    Effective Date: March 28, 1994.
    Amendments Nos.: 186 and 191.
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Licenses.
    Date of Initial Notice in Federal Register: August 5, 1992 (57 FR 
34590). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 28, 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.

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

    Date of Application for Amendments: November 17, 1993.
    Brief Description of Amendments: These amendments revise the 
surveillance requirements to eliminate unnecessary diesel generator 
testing when a diesel generator or an offsite power source becomes 
inoperable. This change reduces the stresses on the diesel generators 
caused by unnecessary testing.
    Date of Issuance: April 5, 1994.
    Effective Date: April 5, 1994.
    Amendments Nos.: 187 and 192.
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: January 5, 1994 (59 FR 
628). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 5, 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.

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

    Date of Application for Amendments: May 25, 1993, as supplemented 
March 11, 1994.
    Brief Description of Amendments: These administrative amendments 
(1) remove references to the Service Platform Hoist, (2) correct a 
typographical error concerning the Emergency Transformer Degraded 
Voltage relay setpoint tolerance, and (3) clarify that the basis for 
recalibration of certain pressure switches is reactor thermal power 
instead of turbine first stage pressure.
    Date of Issuance: April 7, 1994.
    Effective Date: April 7, 1994.
    Amendments Nos.: 188 and 193.
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: July 21, 1993 (58 FR 
39059). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 7, 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.

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

    Date of Application for Amendments: December 21, 1993, as 
supplemented on March 11, 1994.
    Brief Description of Amendments: These amendments revise Technical 
Specification (TS) Table 3.2.F to accurately describe the main stack 
high range and reactor building roof vent high range radiation 
monitors, and delete previously approved Amendment No. 168 for Unit 3. 
Amendment No. 168 was an emergency temporary change which is no longer 
requested.
    Date of Issuance: April 7, 1994.
    Effective Date: April 7, 1994.
    Amendments Nos.: 189 and 194.
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: February 16, 1994 (59 
FR 7697) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 7, 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.

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

    Date of Application for Amendment: December 28, 1993.
    Brief Description of Amendment: The amendment clarifies Limiting 
Condition for Operation (LCO) 3.5.D.4. Amendment No. 179 to the TS 
added LCO 3.5.D.4 to permit hydrostatic and leakage testing at 
temperatures up to 300  deg.F without requiring certain equipment, 
including the automatic depressurization system (ADS), to be operable. 
However, LCO 3.5.D.4 can be mistakenly interpreted to require the ADS 
be operable at temperatures less than 212  deg.F. Requiring the ADS to 
be operable during hydrostatic and leakage testing with temperatures 
below 212  deg.F was clearly not the intent of Amendment No. 179. The 
amendment clarifies LCO 3.5.D.4 to resolve this concern and is 
considered an administrative change.
    Date of Issuance: April 6, 1994.
    Effective Date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10014) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of Application for Amendment: December 29, 1993.
    Brief Description of Amendment: The Technical Specifications (TSs) 
amendment revised Section 3.6.D.4 to eliminate an inconsistency between 
the operability requirements for the reactor coolant system (RCS) 
leakage detection and the specified requirements for monitoring RCS 
leakage. Additionally, the amendment revised the TSs to make numerous 
editorial corrections which are administrative in nature.
    Date of Issuance: April 13, 1994.
    Effective Date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: February 2, 1994 (59 FR 
4945) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of Application for Amendment: August 30, 1993, and supplement 
dated March 21, 1994.
    Brief Description of Amendment: The amendment revises the 
composition of the Station Operations Review Committee (SORC) and 
increases the submittal interval of the Radiological Effluent Release 
Report from semiannually to annually.
    Date of Issuance: April 15, 1994.
    Effective Date: April 15, 1994.
    Amendment No.: 67.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: September 29, 1993 (58 
FR 50973)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 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: April 23, 1993, and supplemented 
by letters dated November 10, 1993 and January 13, 1994.
    Brief Description of Amendment: The amendment lowers the technical 
specification limit for the maximum ultimate heat sink temperature and 
revise the bases for the station service water system.
    Date of Issuance: April 15, 1994.
    Effective Date: April 15, 1994.
    Amendment No.: 68.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register:
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 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 Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of Application for Amendments: December 8, 1993.
    Brief Description of Amendments: These amendments incorporate the 
guidance of NRC Generic Letter 90-06 that addresses power-operated 
relief valve and block valve reliability and additional low-temperature 
overpressure protection for light water reactors.
    Date of Issuance: April 7, 1994.
    Effective Date: April 7, 1994.
    Amendment Nos. 150 and 130.
    Facility Operating License Nos. DPR-70 and DPR-75: These amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: January 19, 1994 (59 FR 
2870) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 7, 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 Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M. 
Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of Application for Amendment: December 9, 1993, as 
supplemented February 23, and April 1, 1994
    Brief Description of Amendment: The amendment modifies Technical 
Specification (TS) 3/4.4.6, Steam Generators, and TS 3/4.4.9, Specific 
Activity, and their associated bases. The steam generator plugging/
repair limit is being modified in the TS to incorporate a 2.0 volt 
steam generator tube support plate interim plugging criteria for Cycle 
13 only. In addition, the TS limit for specific activity of dose 
equivalent I\131\ and its transient dose equivalent I\131\ reactor 
coolant specific activity will be reduced by a factor of 4 in order to 
increase the allowable leakage in the event of a steam line break for 
Cycle 13 only.
    Date of Issuance: April 5, 1994.
    Effective Date: April 5, 1994.
    Amendment No.: 106.
    Facility Operating License No. NPF-2. Amendment revises the 
Technical Specifications.
    Date of Initial Notice in Federal Register: January 19, 1994 (59 FR 
2871) The February 23, 1994, and April 1, 1994, letters provided 
supplemental information and deleted the requested TS upper limit 
bobbin voltage of 5.7 volts for tube plugging that was requested in the 
December 9, 1993, letter and retained the current value of 3.6 volts. 
The February 23 and April 1, 1994, supplements did not change the 
original no significant hazards consideration finding.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 5, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

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

    Date of Amendment Request: May 21, 1993, as supplemented by letter 
dated February 23, 1994.
    Brief Description of Amendment: The amendments change the technical 
specifications by replacing the requirements associated with the 
control room heating and ventilation system with requirements related 
to operation of the control room filtration system and control room air 
conditioning system. The proposed change is consistent with the 
requirements of the Westinghouse Standard Technical Specifications 
(NUREG-1431) issued on September 28, 1992.
    Date of Issuance: April 6, 1994
    Effective Date: April 6, 1994, to be implemented within 30 days of 
issuance.
    Amendment Nos: Unit 1--Amendment No. 23; Unit 2--Amendment No. 9
    Facility Operating License Nos. NPF-87 and NPF-89: Amendments 
revised the Technical Specifications.
    Date of Initial Notice in Federal Register: August 18, 1993 (58 FR 
43933). The February 23, 1994, submittal provided supplemental 
information to the application and did not change the initial no 
significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.

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

    Date of Application for Amendment: September 24, 1993.
    Brief description of amendment: The amendment revises the Technical 
Specifications to extend the reporting period of the Semiannual 
Radioactive Effluent Release Report from semiannually to annually. 
Additionally, the report submission date changes from 60 days after 
January 1 and July 1 of each year to before May 1 of each year. The 
changes to the reporting period and report date are being made to 
implement the August 31, 1992, change to 10 CFR 50.36a. The affected 
Technical Specification Sections are 1.18, 3.11.1.4, 3.11.2.6, 6.9.1.7, 
6.14c, and the Index.
    Date of Issuance: April 14, 1994.
    Effective Date: April 14, 1994.
    Amendment No.:  89.
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10016).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of Application for Amendment: February 1, 1994.
    Brief Description of Amendment: The amendment revises the TS by 
removing the review of the Emergency Plan and its implementing 
procedures from the list of responsibilities of the Plant Operations 
Review Committee (PORC). Guidance for this change was provided in 
Generic Letter 93-07, ``Modification of the Technical Specification 
Administrative Control Requirements for Emergency and Security Plans,'' 
dated December 28, 1993. Several other administrative TS changes were 
also made including removing specific titles from the list of PORC 
members in TS 6.5.a.2 and deleting TS 6.5.b which describes the 
Corporate Support Staff.
    Date of Issuance: April 7, 1994.
    Effective Date: April 7, 1994.
    Amendment No.: 107.
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: March 2, 1994 (59 FR 
10017) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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

    Date of Application for Amendment: May 5, 1993 as supplemented 
March 4, 1994.
    Brief Description of Amendment: The amendment changes the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TS) in response to 
NRC Generic Letter 90-06. This letter deals with Generic Issue 70 and 
Generic Issue 94, which focus on power-operated relief valve and block 
valve reliability and additional low-temperature overpressure 
protection. The amendment revises TS Section 3.1 by adding restrictions 
on the restart of an inactive reactor coolant pump, modifying the 
limiting conditions for operation of the pressurizer power-operated 
relief valves (PORVs) and associated block valves, and adding 
provisions to ensure that adequate low-temperature overpressure 
protection (LTOP) is available. Additionally, this amendment modifies 
the limiting conditions for operation for reactor coolant temperature 
and pressure by adding Figure TS 3.1-4 to define 10 CFR 50 Appendix G 
pressure and temperature limitations for LTOP evaluation through the 
end of operating cycle 20.
    Date of Issuance: April 7, 1994.
    Effective Date: April 7, 1994.
    Amendment No.: 108.
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of Initial Notice in Federal Register: July 21, 1993 (58 FR 
39062) The March 4, 1994, submittal provided additional clarifying 
information and changed the LTOP allowed outage time from 7 days to a 
more conservative 5 days. This modification did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

    Dated at Rockville, Maryland, this 20th day of April 1994.

    For the Nuclear Regulatory Commission.
Gus C. Lainas,
Acting Director, Division of Reactor Projects--I/II, Office of Nuclear 
Reactor Regulation.
[FR Doc. 94-10011 Filed 4-26-94; 8:45 am]
BILLING CODE 7590-01-P