[Federal Register Volume 60, Number 227 (Monday, November 27, 1995)]
[Notices]
[Pages 58394-58416]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-28606]



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NUCLEAR REGULATORY COMMISSION

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from October 28, 1995, through November 9, 1995. 
The last biweekly notice was published on Wednesday, November 8, 1995 
(60 FR 56361).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By December 27, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any 

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

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: October 20, 1995.
    Description of amendment request: The proposed one-time amendment 
would revise the Calvert Cliffs Nuclear Power Plant, Unit No. 1, (CC-1) 
Technical Specifications (TSs) by extending certain 18-month instrument 
surveillance intervals by a maximum of 39 days to March 31, 1996. The 
instruments involved are included in the reactor protective system, 
engineered safety features actuation system, power-operated relief 
valves, low-temperature overpressure protection system, remote shutdown 
instruments, post-accident monitoring, radiation monitoring, and 
containment sump level instruments.
    The Commission issued Amendment No. 208 to Facility Operating 
License No. DRP-53 and Amendment No. 186 to Facility Operating License 
No. DRP-69 for the CC-1/2, respectively. The amendments permanently 
extended the surveillance intervals for the instruments described above 
from 18 months to 24 months after a specified number of the instruments 
had been replaced. The amendments were effective immediately and to be 
implemented on CC-2 within 30 days, but not implemented on CC-1 until 
its restart after the spring 1996 refueling outage. All of the 
instruments identified for replacement on CC-2 have been replaced, but 
those identified for replacement on CC-1 have not been replaced, thus, 
the reason for the later implementation date. The proposed one-time 
amendment is needed prior to Amendment No. 208 being implemented 
because of a change in the refueling schedule. The licensee has 
provided technical justification to allow operation for an additional 
short-time period of up to a maximum of 39 days.
    CC-1 was initially scheduled to begin its refueling outage on 
February 16, 1996, which would have been within the time frame 
necessary to perform the required 18-month instrument surveillances 
currently required for the instruments identified above. The licensee 
has recently rescheduled the refueling outage for CC-1 to start March 
15, 1996, several months after the initial amendment request and after 
consultation with the Pennsylvania-New Jersey-Maryland power pool. The 
revised schedule will allow the maximum use of the available fuel in 
the CC-1 reactor core and will also allow the unit to operate for an 
additional period of about 1 month during a period of potentially high 
power demand. In addition, the delay will allow more time to plan and 
prepare for the upcoming refueling outage. Performing the required 
instrument surveillances at power would present an unwarranted 
personnel safety risk and, in some cases, the surveillances cannot be 
done during power operation because they would cause a unit trip. This 
proposed one-time amendment will be superseded by Amendment No. 208 
when it is implemented.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed one-time change would extend 18-month instrument 
surveillance intervals by a maximum of 39 days to March 31, 1996, 
for specific Reactor Protective System (RPS), Engineered Safety 
Features Actuation System (ESFAS), Power-Operated Relief Valve, Low 
Temperature Overpressure Protection (LTOP), Remote Shutdown, Post-
Accident Monitoring (PAM), Radiation Monitoring, and Containment 
Sump Level instruments.
    The purpose of the RPS is to effect a rapid reactor shutdown if 
any one or a combination of conditions deviates from a pre-selected 
operating range. The system functions to protect the core and the 
Reactor Coolant System (RCS) pressure boundary. The purpose of the 
ESFAS is to actuate equipment which protects the public and plant 
personnel from the accidental release of radioactive fission 
products if an accident occurs, including a loss-of-coolant 
accident, main steam line break, or loss of feedwater event. The 
safety features function to localize, control, mitigate, and 
terminate such incidents in order to minimize radiation exposure to 
the general public. The PAM instruments provide the Control Room 
operators with primary information necessary to take manual actions, 
as necessary, in response to design basis events, and to verify 
proper system response to plant conditions and operator actions. The 
purpose of the Remote Shutdown System is to provide plant parameter 
indications to operators on a Remote Shutdown Panel to be used while 
placing and maintaining the plant in a safe shutdown condition in 
the event the Control Room is uninhabitable. The indications are 
used to verify proper system response to plant conditions and 
operator actions. The LTOP System protects against RCS 
overpressurization at low temperatures 

[[Page 58397]]
by a combination of administrative controls and hardware. Power-
Operated Relief Valves are set to lift before pressurizer safety 
valves, and subsequently reseat to minimize the release of reactor 
coolant from the RCS. The Containment Sump High Level Alarm System 
provides an alarm in the Control Room to provide one of the 
available indications of excessive RCS leakage during normal plant 
operation. The Containment Area High Range Radiation Monitoring 
System provides an indication of high radiation levels in 
containment.
    Failure of any of these systems is not an initiator for any 
previously evaluated accident. Therefore, the proposed change would 
not involve an increase in the probability of an accident previously 
evaluated.
    Surveillance and maintenance history has demonstrated good 
capability for identifying adverse operation by individual 
instruments. Baltimore Gas and Electric Company has the capability 
to respond to an inoperable instrument by following the Technical 
Specification Actions for an inoperable instrument or by performing 
a channel calibration with the Unit at full power. However, 
calibration of all the instruments at power is not desirable because 
of personnel safety, personnel radiation protection goals, and plant 
reliability concerns.
    These factors provide assurance that the requested surveillance 
extension will not adversely affect our ability to detect 
degradation of the instruments. Also, either analysis is available 
to show the instruments will operate properly during the requested 
surveillance extension, or the surveillance program has shown that 
problems will be identified and addressed appropriately. Therefore, 
these channels will be able to perform the functions assumed in the 
safety analysis, and there is no significant increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed Technical Specification changes do not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    This requested increase in surveillance interval for RPS, ESFAS, 
Power-Operated Relief Valve, LTOP, Remote Shutdown, PAM, Radiation 
Monitoring, and Containment Sump Level instrument surveillances does 
not involve a significant change in the design or operation of the 
plant. No plant hardware is being modified as part of the proposed 
change. The proposed change also does not involve any new or unusual 
actions by plant operators. Therefore, this change would not create 
the possibility of a new or different type of accident from any 
accident previously evaluated.
    3. Does operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    The RPS, ESFAS, Power-Operated Relief Valve, LTOP, Remote 
Shutdown, PAM, Radiation Monitoring, and Containment Sump Level 
instruments are designed to provide actuation signals and/or 
indications to ensure appropriate action is taken in response to 
design basis accidents. Channel checks, channel functional tests and 
routine comparison of the redundant and independent parameter 
indications provides a reliable indication of instrument operation. 
Also, either analysis is available to show the instruments will 
operate properly during the requested surveillance extension, or 
instrument surveillance program has shown that problems will be 
identified and addressed appropriately. During the requested 
extension, these systems will be available to perform the functions 
assumed in the Safety Analysis. Surveillance and maintenance history 
have demonstrated good capability for identifying adverse operation 
by individual instruments. Baltimore Gas and Electric Company has 
the capability to respond to such adverse operation, including 
performing channel calibrations at power. However, such work on all 
the instruments is not desirable because of personnel safety, 
personnel radiation protection goals, and plant reliability 
concerns. Extending the surveillance interval provides additional 
possibility for instrument components to malfunction by means such 
as drift or instrument failure, which could allow plant parameters 
to exceed design bases assumptions. We have determined that the 
effect of the surveillance interval extension on safety is small, 
and operation of the instruments in the extended interval would not 
invalidate any assumption in the plant licensing basis.

    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: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

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

    Date of amendment request: October 3, 1995.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) for both stations to 
implement 10 of the line item TS improvements recommended in Generic 
Letter (GL) 93-05, ``Line-Item Technical Specifications Improvements to 
Reduce Surveillance Requirements for Testing During Power Operation,'' 
dated September 27, 1993. The proposed changes also include editorial 
changes on the affected TS pages.
    The proposed changes from GL 93-05 are the following: (1) TS 
4.1.3.1.2 (GL 93-05, Item 4.2), extending the interval for checking the 
operability of each full-length rod not fully inserted in the core from 
31 days to 92 days; (2) Table 4.3-3 (GL 93-05, Item 5.14), extending 
the interval for the digital channel operational test for radiation 
monitoring instrumentation in the table from monthly to quarterly; (3) 
TS 4.4.3.2 (GL 93-05, Item 6.6), extending the interval between current 
tests of the required groups of pressurizer heaters from 92 days to 
each refueling outage; (4) TS 4.4.6.2.2.b (GL 93-05, Item 6.1), 
extending the time the plant may be in cold shutdown before pressure 
isolation valve testing is required, prior to entry into Operational 
Mode 2, from 72 hours to 7 days; (5) TS 4.5.1.1.b (GL 93-05, Item 7.1), 
revising the requirement to verify the boron concentration in an 
accumulator within 6 hours of any volume increase to the accumulator 
(greater than or equal to 70 gallons) so that the verification is not 
required when the volume increase is from the refueling water storage 
tank (RWST) and the RWST has not been diluted since verifying that the 
boron concentration of the RWST is within the concentration limits for 
the accumulators; (6) TS 4.6.2.1 (GL 93-05, Item 8.1), extending the 
interval between tests to verify each containment spray nozzle is 
unobstructed from 5 years to 10 years; (7) TS 4.6.4.1 (GL 93-05, Item 
5.4), extending the interval for testing each hydrogen monitor for 
combustible gas control from 31 days to 92 days for the analog channel 
operational test, and from 92 days to each refueling outage for channel 
calibration; (8) TS 4.6.4.2 (GL 93-05, Item 8.5), extending the 
interval between tests to demonstrate operability of the hydrogen 
recombiner system from 6 months to once each refueling outage; (9) TS 
4.7.1.2.1.a (GL 93-05, Item 9.1), extending the interval between tests 
of the auxiliary feedwater pumps from 31 days to 92 days on a staggered 
test basis; and (10) TS 4.11.2.6 (GL 93-05, Item 13), extending the 
interval for determining the quantity of radioactivity contained in 
each gas decay tank, when radioactivity is being added to the tanks, 
from 24 hours to 7 days, with the 24-hour frequency maintained during 
the primary coolant degassing operation. The editorial changes are the 
following: (1) TS 4.4.6.2.1.c, changes the word ``from'' to the word 
``to,'' (2) TS 4.5.1.1.c, the 

[[Page 58398]]
change clarifies that the motor control center compartment is for each 
accumulator isolation valve, (3) TS 4.5.1.2, deletes the footnote 
because the operating cycle in the footnote is over for each unit, and 
(4) TS 4.7.1.2.1.a.2 and 4.7.1.2.1.c, renumbers and rephrases (only TS 
4.7.1.2.1.a.2) other surveillance requirements for the auxiliary 
feedwater pumps because of the proposed change to TS 4.7.1.2.1.a to 
implement GL 93-05, Item 9.1.
    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:

    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes are consistent with GL 93-05 and NUREG-1366 
[''Improvements to Technical Specifications Surveillance 
Requirements,'' December 1992. In GL 93-05, the staff stated that it 
concluded, in performing the study documented in NUREG-1366, that 
safety can be improved, equipment degradation decreased, and an 
unnecessary burden on licensee personnel eliminated by reducing the 
frequency of certain testing required in the Technical 
Specifications during power operation]. The changes eliminate 
testing that is likely to cause transients or excessive wear of 
equipment. An evaluation of these changes indicates that there will 
be a benefit to plant safety. The evaluation, documented in NUREG-
1366, considered (1) unavailability of safety equipment due to 
testing, (2) initiation of significant transients due to testing, 
(3) actuation of engineered safety features that unnecessarily cycle 
safety equipment, (4) importance to safety of that system or 
component, (5) failure rate of that system or component, and (6) 
effectiveness of the test in discovering the failure.
    As a result of the decrease in the testing frequencies, the risk 
of testing causing a transient and equipment degradation will be 
decreased, and the reliability of the equipment will not be 
significantly decreased.
    The initial conditions and methodologies used in the accident 
analyses remain unchanged. The proposed changes do not change or 
alter the design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, accident 
analyses results are not impacted. Appropriate testing will continue 
to assure that equipment and systems will be capable of performing 
the intended function.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or 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 either modify allowable intervals between 
certain surveillance tests, delete surveillance requirements, or 
alter an action statement with regard to the required testing. The 
proposed changes do not affect the design or operation of any 
system, structure, or component in the plant. The safety functions 
of the related structures, systems, or components are not changed in 
any manner, nor is the reliability of any structure, system, or 
component reduced by the revised surveillance or testing 
requirements.
    Appropriate testing will continue to assure that the system is 
capable of performing its intended function. The changes do not 
affect the manner by which the facility is operated and do not 
change any facility design feature, structure, system, or component. 
No new or different type of equipment will be installed. Since there 
is no change to the facility or operating procedures, and the safety 
functions and reliability of structures, systems, or components are 
not affected, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    All of the proposed technical specification changes are 
compatible with plant operating experience and are consistent with 
the guidance provided in GL 93-05 and NUREG-1366. The changes 
eliminate unnecessary testing that increases the risk of transients 
and equipment degradation. There is no impact on safety limits or 
limiting safety system settings.
    The remaining proposed changes are administrative in nature and 
have no impact on the margin of safety of any technical 
specification. They do not affect any plant safety parameters or 
setpoints.

    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 District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington 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 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: October 2, 1995
    Description of amendment request: The proposed amendments would 
revise Section 3.4.2 to change the safety/relief valve (SRV) safety 
function lift setting tolerances from +1%, -3% to plus or minus 3% and 
include as-left SRV safety function lift setting tolerances of plus or 
minus 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 staff has reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below.
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The probability of an accident previously evaluated will not 
increase as a result of this change, because the only changes are the 
tolerances for the SRV opening setpoints and the speed of the reactor 
core isolation cooling system (RCIC) turbine and pump. Changing the 
maximum allowable opening setpoint for the SRVs does not cause any 
accident previously evaluated to occur, or degrade valve or system 
performance in any way so as to cause an accident to occur with an 
increased frequency. In addition, the increased speed of the RCIC 
turbine and pump are within the design limits of the system. RCIC 
operability and failure probabilities are not impacted by this change.
    The consequences of an ASME Overpressurization Event are not 
significantly increased and do not exceed the previously accepted 
licensing criteria for this event. General Electric (GE) has calculated 
the revised peak vessel pressure for LaSalle Station to be 1341 psig, 
which is well below the 1375 psig criterion of the ASME Code for upset 
conditions, referenced in Section 5.2.2, Overpressurization Protection, 
of the Updated Final Safety Analysis Report (UFSAR), and NUREG-0519 
(Safety Evaluation Report related to the operation of LaSalle County 
Station, Units 1 and 2, March 1981), and Section 15.2-4, Closure of 
Main Steam Isolation Valves (BWR) of NUREG-0800 (Standard Review Plan).
    GE has also performed an analysis of the limiting Anticipated 
Transient Without Scram (ATWS) event, which is the Main Steam Isolation 
Valve (MSIV) Closure Event. This analysis calculated the peak vessel 
pressure to be 1457 psig, which is sufficiently below the 1500 psig 
criterion of the ASME Code for emergency conditions.
    Per NUREG-0519, listed above, Section 5.4.1, and Technical 
Specification 4.7.3.b, the RCIC pump is required to develop flow 
greater than or 

[[Page 58399]]
equal to 600 gpm in the test flow path with a system head corresponding 
to reactor vessel operating pressure when steam is supplied to the 
turbine at 1000 +20, -80 psig. Increasing the turbine and pump speed 
ensures these criteria will still be met and the consequences of an 
accident will not increase.
    Therefore, there is not a significant increase in the consequences 
of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The only physical changes are to increase the allowable tolerances 
for SRV opening setpoints and to increase the RCIC pump and turbine 
speeds. These changes do not result in any changed component 
interactions. The SRVs and RCIC will still provide the functions for 
which they were designed. Since all of the other systems evaluated will 
continue to function as intended, 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 
the margin of safety.
    While the calculated peak vessel pressures for the ASME 
Overpressurization Event and the MSIV closure ATWS Event are larger 
than that previously calculated without the proposed setpoint tolerance 
increases, the new peak pressures remain sufficiently below the 
respective licensing acceptance limits associated with these events. In 
addition, the actual L1C8 reload analysis of the ASME 
Overpressurization Event will be verified to be within the licensing 
acceptance limit for that event prior to Unit 1 Cycle 8 startup, as 
required in the normal reload 10 CFR 50.59 process. These licensing 
acceptance limits have been previously evaluated as providing a 
sufficient margin of safety. For other accidents and transients, the 
increased setpoint tolerances have a negligible effect on the results, 
so the margin of safety is preserved.
    The staff has reviewed the amendment request and the licensee's no 
significant hazards consideration determination. Based on the review 
and the above discussions, the staff proposes to determine that the 
proposed changes do not involve a significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: October 17, 1995.
    Description of amendment request: The proposed amendment would 
modify the Palisades Facility Operating License to reference 10 CFR 
Part 40, allow the use of source materials as reactor fuel, delete 
references to specific amendments and specific revisions in the listed 
titles of the Physical Security Plan Suitability Training and 
Qualification Plan and the Safeguards Contingency Plan, delete 
paragraph 2.F on reporting requirements, and make minor editorial 
changes. In addition, the Technical Specifications (TS) would be 
modified as follows: (1) TS 3.1.2 would be modified to change the 
pressurizer cooldown limit from 100  deg.F to 200  deg.F/hour; (2) the 
shield cooling system requirements would be relocated to the Palisades 
Final Safety Analysis Report (FSAR); (3) several minor editorial 
changes to various sections of the TS are proposed; and (4) revisions 
to several TS bases pages are proposed.
    Basis for proposed no significant hazards consideration 
determination As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

Administrative Changes

    Since these changes have no effect on the physical plant or its 
operation, they cannot involve a significant increase in the 
probability or consequences of an accident previously evaluated, 
create the possibility of a new or different kind of accident from 
any previously evaluated, or involve a significant reduction in a 
margin of safety.

Technical Changes

    The following evaluation supports the finding that operation of 
the facility in accordance with the two non-administrative changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Use of Source Material as reactor fuel: The use of depleted or 
natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in 
addition to the currently allowed ``slightly enriched uranium'' 
would not affect the physical plant or its operation in any way 
which could increase the probability of any previously evaluated 
accident. Its use would not introduce any new kind or additional 
amount of fission product material. Therefore, use of source 
material as reactor fuel would not affect the consequences of an 
accident previously evaluated.
    Restoration of the Pressurizer Cooldown Rate Limit: The 
Palisades Technical Specifications contain a single limit, item 
3.1.2 b, for both heatup and cooldown rates for the pressurizer. The 
October 5, 1994 change request proposed changing that limit from 
200 deg.F/hour to 100 deg.F/hour solely due to its inconsistency 
with the pressurizer design analysis. Fatigue calculations in the 
pressurizer design analysis assumed a heatup rate of 100 deg.F/hour 
and a cooldown rate of 200 deg.F/hour. Until issuance of Amendment 
163, the Technical specifications contained a single limit for both 
heatup and cooldown rates of 200 deg.F/hour. Although the installed 
equipment is not capable of exceeding the 100 deg.F/hour heatup 
limit, the October 5, 1994 change request proposed a revised limit 
to assure that the Technical Specification limit was not less 
restrictive than the design analysis. The higher pressurizer 
cooldown rate does not affect the results of our analyses which 
determined the PCS Pressure-Temperature limits or the [Loss of 
Temperature Overpressurization] LTOP setting requirements of the 
Technical Specifications.
    When the change was proposed, it was not realized that the more 
limiting cooldown rate might adversely, and unnecessarily, affect 
plant operation. This proposed change to the Technical 
Specifications would separate the limits for heatup rate and 
cooldown rate, returning the specified cooldown rate to the original 
value which was consistent with plant design. The current heatup 
rate limit, which is also consistent with the design, would be 
retained. The proposed pressurizer cooldown rate will allow 
depressurizing of the primary coolant system [PCS] and flooding the 
pressurizer steam space without undue restriction. The more rapid 
depressurization would be important in the event of a steam 
generator tube rupture.
    Therefore, operation of the facility in accordance with the 
proposed change to the Technical Specifications would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Use of Source Material as reactor fuel: The use of depleted or 
natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in 
addition to the currently allowed ``slightly enriched uranium'' 
would not affect the design (other than the fuel enrichment), 
configuration, or operation of the plant. Therefore this change 
cannot create the possibility of a new or different kind of accident 
from any previously evaluated.
    Restoration of the Pressurizer Cooldown Rate Limit: The proposed 
change to the Technical Specifications would bring the plant within 
the assumptions of the design documents for the pressurizer and in 
line with the Accident analysis for the rapid reduction of the 
primary coolant system pressure. With the lower rate specified in 
the present technical specification, the depressurization of the PCS 
will be delayed to maintain the lower pressurizer cooldown rate.
    Therefore, operation of the facility in accordance with the 
proposed change to the 

[[Page 58400]]
Technical Specifications would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Use of Source Material as reactor fuel: The use of depleted or 
natural uranium, defined as ``Source Material'' by 10 CFR 40.4, in 
addition to the currently allowed ``slightly enriched uranium'' 
would not affect the Safety Limits, Limiting Conditions for 
Operation or other operating limits, or the safety analyses which 
they support. Therefore, the margin of safety is unaffected.
    Restoration of the Pressurizer Cooldown Rate Limit: The proposed 
change to the Technical Specifications would bring the plant in line 
with the design analysis. This will not reduce the margin of safety 
since the higher rate is the basis for the present margin of safety.
    Therefore, the proposed change to the Technical Specifications 
would not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Brian E. Holian, Acting.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of amendment request: September 20, 1995.
    Description of amendment request: The proposed amendment would 
allow a one-time extension of the 18-month surveillance intervals 
contained in the Technical Specifications (TS) related to system 
testing, instrumentation calibration, component inspection, component 
testing, response time testing and logic system functional tests for 
various systems, components and instrumentation.
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS changes involve a one-time only change in the 
surveillance testing intervals to facilitate a one-time only change 
in the Fermi 2 operating cycle. The proposed TS changes do not 
physically impact the plant nor do they impact any design or 
functional requirements of the associated systems. That is, the 
proposed TS changes do not significantly degrade the performance or 
increase the challenges of any safety systems assumed to function in 
the accident analysis. The proposed TS changes affect only the 
frequency of the surveillance requirements and do not impact the TS 
surveillance requirements themselves. In addition, the proposed TS 
changes do not introduce any new accident initiators since no 
accidents previously evaluated have as their initiators anything 
related to the change in the frequency of surveillance testing. 
Also, the proposed TS changes do not significantly affect the 
availability of equipment or systems required to mitigate the 
consequences of an accident because of other, more frequent testing 
or the availability of redundant systems or equipment. Furthermore, 
a historical review of surveillance test results support the above 
conclusions. Therefore, the proposed TS changes do not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes involve a one-time only change in the 
surveillance testing intervals to facilitate the one-time only 
change in the Fermi 2 operating cycle. The propose TS changes do not 
introduce any failure mechanisms of a different type than those 
previously evaluated since there are no physical changes being made 
to the facility. In addition, the surveillance test requirements 
themselves will remain unchanged. Therefore, the proposed TS changes 
do not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Although the proposed TS changes will result in an increase in 
the interval between some surveillance tests, the impact, if any, on 
system availability is small based on other, more frequent testing 
or redundant systems or equipment, and there is no evidence of any 
time dependent failures that would impact the availability of the 
systems. Therefore, the assumptions in the licensing basis are not 
impacted, and the proposed TS changes do not significantly reduce a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Brian E. Holian, Acting.

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

    Date of amendment request: August 8, 1995.
    Description of amendment request: The amendments would revise 
Technical Specification Section 3/4.4.8, Table 4.4-4, Table Notations, 
to allow the reactor coolant system gross specific activity measurement 
method to be changed from the current degassed method to a non-
degassed, or pressurized dilution, method.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1

    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The amendments will have no effect on the probability of 
the occurrence of any accident. It has been demonstrated that the 
results obtained by the pressurized dilution technique are 
statistically similar to results obtained by the degassed technique. 
Therefore, implemention of the new method will have no effect 
insofar as the accuracy of the NC [reactor coolant system] system 
specific activity determination is concerned. Therefore, there will 
be no effect upon any accident dose consequences.

Criterion 2

    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No accident causal mechanisms will be affected by 
installation of the sampling equipment required by the pressurized 
dilution technique. Operation of the NC system itself will not be 
affected by the proposed change in sampling technique. All procedure 
changes required for implementation of the new sampling method will 
be made according to the provisions of 10 CFR 50.59. No impact on 
other areas of plant operations will be generated as a result of the 
new sampling method.

Criterion 3

    The requested amendments will not involve a significant 
reduction in a margin of safety. No impact on any safety limits will 
result from the change in sample method from the degassed technique 
to the pressurized dilution technique. Several benefits will result 
from the change, 

[[Page 58401]]
including fewer opportunities for valve mispositionings to occur, as 
well as reduced radiation exposure to Chemistry technicians. The 
proposed amendment is consistent with a similar amendment approved 
by the NRC for McGuire Nuclear Station (Amendment Nos. 66 and 47 for 
McGuire Units 1 and 2, respectively).
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: November 7, 1995.
    Description of amendment request: The proposed change would revise 
Technical Specification 3/4.5.1 SAFETY INJECTION TANKS (SITs) by 
increasing the specified range associated with SIT water level and 
nitrogen cover pressure.
    The current limiting conditions for operation (LCO) for the SIT 
requires that four SITs be operable with a water volume in the range of 
1679 cubic feet (78%) to 1807 cubic feet (83.8%) and a nitrogen cover 
pressure between 600 psig to 625 psig. The proposed change requests an 
expanded range of 925.6 cubic feet (40%) to 1807 cubic feet (83.8%) for 
SIT level and 600 psig to 670 psig for SIT pressure indicators.
    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:

    Operation of the facility in accordance with this change does 
not involve an increase in the probability of any accident. The SITs 
are used to mitigate the consequences of an accident and are not 
accident initiators.
    The proposed change would actually decrease the consequence of 
events such as LOCA [loss of coolant accident] which would result in 
rapid RCS [reactor coolant system] depressurization.
    By reducing SIT level, the initial nitrogen gas volume is 
increased which results in an increase in the SIT flow rate into the 
RCS for a given RCS pressure transient. This decreases the time 
required to fill the reactor vessel lower plenum after the end of 
blowdown. During refill, fuel cladding temperature increases rapidly 
due to insufficient cooling which is provided solely by rod to rod 
thermal radiation. Decreasing the refill time therefore, results in 
lower cladding temperature at the start of core reflood which 
results in lower Peak Cladding Temperature (PCT) during reflood.
    Increasing the nitrogen cover pressure would also result in 
increased SIT flow rate and would be beneficial as described above.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequence of any accident.
    The proposed change will not create any new system connections 
or interactions. Thus, no new modes of failure are introduced. The 
increased range for SIT pressure and level is actually beneficial in 
maintaining lower PCT following a LOCA.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The impact of the proposed changes on the Waterford 3 FSAR 
[Final Safety Analysis Report] analyses have been evaluated. The AOR 
[Analysis of Record] shows that PCT and maximum cladding oxidation 
would increase slightly as a result of this change. However, they 
both remain below the acceptance criteria values of 2200 degrees 
fahrenhit and 17% for PCT and maximum cladding oxidation, 
respectively. The system capabilities to mitigate the consequences 
of accidents will be the same as they were prior to these changes.
    Therefore, the proposed changes do[es] 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: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street NW, Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: August 10, 1995
    Description of amendment request: This amendment would incorporate 
certain improvements into the Three Mile Island, Unit 1 Technical 
Specifications consistent with the Standard Technical Specifications 
for Babcock and Wilcox plants. The requested changes would affect the 
reactor building isolation instrumentation, sampling frequency for the 
sodium hydroxide tank, and the surveillance requirements for the plant 
vital bus batteries.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    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 proposed amendment involves changes to the 
TMI-1 Technical Specifications [TS] which are consistent with the 
[Babcock & Wilcox] B&W Standard Technical Specifications ([R]STS), 
NUREG-1430. This change does not involve any change to system or 
equipment configuration. The proposed amendment revises certain 
surveillance requirements, or extends certain surveillance 
intervals. The reliability of systems and components relied upon to 
prevent or mitigate the consequences of accidents previously 
evaluated is not degraded by the proposed changes. Therefore, this 
change does not involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The change 
only involves changes to surveillance requirements that are 
consistent with RSTS or deletion of requirements which are not 
appropriate for TS. No new failure modes are created and thus the 
changes are bounded by accidents previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. These proposed changes involve deletions of requirements or 
changes in surveillance requirements consistent with the B&W RSTS. 
No operating limits are affected and no reduction in the margin of 
safety is involved.

    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: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

[[Page 58402]]

    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London, Connecticut

    Date of amendment request: October 24, 1995.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Surveillance Requirement of 
Section 4.4.5.1, ``Steam Generators'' and the Bases for Section 3/
4.4.5, ``Steam Generators.'' Typographical errors in Section 
4.4.5.1.3.c.1 and Table 4.4-6 are also proposed to be corrected. The 
proposed amendment would defer the next required surveillance to 
inspect steam generator tubes from October 20, 1996, to the next 
refueling outage or no later than October 20, 1997, whichever is 
earlier.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    Pursuant to 10 CFR 50.92, NNECO [the licensee] has reviewed the 
proposed one-time change to extend the maximum allowable inspection 
interval for steam generator tubes from 24 months to 36 months. 
NNECO concludes that these changes do not involve a significant 
hazards consideration since the proposed change satisfies the 
criteria in 10 CFR 50.92(c). That is, the proposed changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    This change involves one-time deferment of the eddy current 
inspection of the steam generator tubes until the end of the next 
refueling outage following the thirteenth fuel cycle, but no longer 
than 12 months beyond the original due date for the inspection. The 
steam generator tubes have only been exposed to one operating cycle 
and are made of thermally treated Alloy 690, one of the most 
corrosion resistant material currently used in recirculating steam 
generators. Following the first full fuel cycle of operation, the 
steam generator tube inspection found the tubes to be in excellent 
condition (i.e., no repairs were required and there was no evidence 
of an active degradation mechanism). Accordingly, no significant 
tube degradation is expected by the end of the thirteenth fuel 
cycle. Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    This one-time change, allowing the steam generator tubes to be 
examined at the end of the refueling outage following Cycle 13 does 
not alter the physical design, configuration, or method of operation 
of the plant. The extension of the inspection interval is not 
expected to result in significant steam generator tube degradation. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    Steam generator tube degradation occurs primarily during 
operation. The change to extend the maximum allowable inspection 
interval for steam generator tubes from 24 months to 36 months will 
not significantly increase the total operating time during Cycle 13 
(the plant was in an outage for at least 10 months of the 12 month 
extension). Therefore, there is no significant effect on the extent 
and severity of tube degradation. The improved corrosion resistance 
of the steam generators tubes (thermally treated Alloy 690) 
minimizes the threat of primary- and secondary-side corrosion. No 
indications of corrosion have been identified in inspections 
performed so far. Based on our assessment of the inspection data and 
corrosion potential, all tubes are expected to be within the 
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR Steam 
Generator Tubes,'' limits by the end of Cycle 13. Also, correction 
of the typographical errors will improve the fidelity of the 
specification. Therefore, this change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: May 1, June 14 and 29, July 14, 17, 18, 
and 26, 1995 with supplemental information provided by letter dated 
October 20, 1995.
    Description of amendment request: Each proposed amendment would 
change the surveillance requirement frequency from the current once per 
18-month interval to once per 24-month which is the current length of a 
Millstone Unit 3 refueling cycle. The changes pertain to the following 
equipment:
    May 1, 1995, Flow Paths--Operating; Position Indication System; Rod 
Drop Time; Seismic Monitoring System; Loose Part Detection System; 
Quench Spray System; Containment Recirculation Spray System; 
Containment Isolation Valves. This notice supersedes the notice 
published in the Federal Register on June 6, 1995 (60 FR 29882) 
relating to containment isolation valves.
    May 1, 1995, Steam Generator Tube Inspections; 10CFR50, Appendix J, 
Type B and Type C Tests.
    June 14, 1995, AC Sources Operating; DC Sources Operating; 
Containment Penetration Conductor Overcurrent Protective Devices; 
Motor-Operated Valves Thermal Overload Protection.
    June 29, 1995, Electric Hydrogen Recombiners; Auxiliary Feedwater 
System; Reactor Plant Component Cooling Water System; Service Water 
System; Snubbers.
    July 14, 1995, ECCS Subsystems--Tavg Greater Than or Equal to 350 
deg.F; pH Trisodium Phosphate Storage Baskets.
    July 17, 1995, Supplementary Leak Collection and Release System; 
Control Room Emergency Ventilation System; Control Room Envelope 
Pressurization System; Auxiliary Building Filter System; Fuel Building 
Exhaust Filter System.
    July 18, 1995, Reactor Coolant System.
    July 26, 1995; Reactor Trip System Instrumentation; ESFAS 
Instrumentation; Remote Shutdown Instrumentation; Accident Monitoring 
Instrumentation; RCS Total Flow Rate; Process and Radiation Monitoring 
Instrumentation.
    In addition, the specifications are changed from a five-column to a 
one-column format.
    Basis for proposed no significant hazards consideration 
determination: The Commission has made a proposed determination that 
the amendment request involves 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 

[[Page 58403]]
(3) involve a significant reduction in a 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. 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 changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to surveillance requirements of the Millstone 
Unit No. 3 Technical Specifications extend the frequency for checking 
the operability of the affected components/equipment. The proposal 
would extend the frequency from at least once per 18 months to at least 
once each refueling interval (i.e., nominal 24-months).
    Changing the frequency of surveillance requirements from at least 
once per 18 months to at least once each refueling interval does not 
change the basis for the frequency. The frequency was chosen because of 
the need to perform this verification under the conditions that apply 
during a plant outage, and to avoid the potential of an unplanned 
transient if the surveillances were conducted with the plant at power.
    The proposed changes do not alter the intent or method by which the 
surveillances are conducted, do not involve any physical changes to the 
plant, do not alter the way any structure, system, or component 
functions, and do not modify the manner in which the plant is operated. 
As such, the proposed changes in the frequency of surveillance 
requirements will not degrade the ability of the equipment/components 
to perform its safety function.
    Additional assurance of the operability of the components/equipment 
is provided by additional surveillance requirements (e.g., monthly or 
quarterly surveillances).
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
surveillance requirements. This evaluation included a review of 
surveillance results, preventive maintenance records, and the frequency 
and type of corrective maintenance. It concluded that there is no 
indication that the proposed extension could cause deterioration in the 
condition or performance of any of the subject components.
    In addition to the substantive changes, there are format changes 
which are merely editorial and because format changes produce no 
physical change they do not influence the probability or consequences 
of accidents.
    Since the proposed changes only affect the surveillance frequency 
for safety systems that are used to mitigate accidents, the changes 
cannot affect the probability of any previously analyzed accident. 
While the proposed changes can lengthen the intervals between 
surveillances, the increases in intervals has been evaluated and it is 
concluded that there is no significant impact on the reliability or 
availability of the safety system and consequently, there is no impact 
on the consequences on any analyzed accident.
    2. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes to surveillance requirements of the Millstone 
Unit No. 3 Technical Specifications extend the frequency for verifying 
the operability of the affected components/equipment. The proposal 
would extend the frequency from at least once per 18 months to at least 
once each refueling interval (nominal 24 months).
    Changing the frequency of surveillance requirements from at least 
once per 18 months to at least once each refueling interval does not 
change the basis for the frequency. The frequency was chosen because of 
the need to perform this verification under the conditions that apply 
during a plant outage, and to avoid the potential of an unplanned 
transient if the surveillances were conducted with the plant at power.
    In addition to the substantive changes, there are format changes 
which are merely editorial and because format changes produce no 
physical change they do not influence the probability of new or 
different types of accidents.
    The proposed changes do not alter the intent or method by which the 
surveillances are conducted, do not involve any physical changes to the 
plant, do not alter the way any structure, system, or component 
functions, and do not modify the manner in which the plant is operated. 
As such, the proposed changes cannot create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The changes do not involve a significant reduction in a margin 
of safety.
    The proposed changes to surveillance requirements of the Millstone 
Unit No. 3 Technical Specifications extend the frequency for verifying 
the operability of the components/equipment. The proposal would extend 
the frequency from at least once per 18-months to at least once each 
refueling interval (24-months).
    In addition to the substantive changes, there are format changes 
which are merely editorial and because format changes produce no 
physical change they do not influence the margin of safety.
    The proposed changes to surveillance frequency are still consistent 
with the basis for the frequency, and the intent or method of 
performing the surveillance is unchanged. Further, the current 
inservice testing requirements and the previous history of reliability 
of the system provides assurance that the changes will not affect the 
reliability of the auxiliary feedwater system. Thus, it is concluded 
that there is no impact on the 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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment requests: September 29, 1995.
    Description of amendment requests: The amendments would add a one-
time footnote to the Technical Specifications regarding the emergency 
diesel generator diesel fuel oil storage and transfer system to permit 
the existing storage tanks to be replaced with double walled tanks and 
piping that comply with new California regulations.
    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.
    Neither the emergency diesel generators (EDGs) nor the diesel 
fuel oil (DFO) storage and transfer system is an accident initiator. 
When performing the modifications to the 

[[Page 58404]]
DFO storage tanks and transfer piping, administrative compensatory 
measures will be taken to reduce the potential challenge to the EDGs 
and to verify the operability of the DFO transfer system. A 
probabilistic risk assessment (PRA) was performed and demonstrates 
that the change in core damage frequency associated with taking each 
DFO storage tank and its associated suction transfer piping out of 
service for 60 days (total of 120 days for both trains) is not 
significant considering the compensatory measures which will be 
taken during the tank replacement period.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Neither the EDGs nor the DFO storage and transfer system is an 
accident initiator. Temporary DFO storage will be onsite during tank 
replacement. The fire protection guidelines in Appendix 9.5B of the 
Updated Final Safety Analysis Report will be complied with in order 
to ensure temporary DFO storage without risk to plant systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes considering implementation of the 
compensatory measures has been shown to not impair safe operation of 
the plant. Having one DFO storage tank and associated piping out of 
service does not reduce the margin of safety since temporary storage 
of DFO will be maintained onsite and administrative compensatory 
measures will be taken to minimize the potential impact of this 
condition. Additionally, delivery of DFO to the site is available 
within 24 hours if needed.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

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

    Date of amendment requests: October 4, 1995.
    Description of amendment requests: The amendments would relocate 
the requirements in ten sub-sections of the Technical Specifications to 
licensee controlled documents in accordance with the guidance in the 
Commission's Final Policy Statement and the Commission's revisions to 
10 CFR 50.36 (60 FR 36959, July 19, 1995) on the content of Technical 
Specifications and the Standard Technical Specifications, Westinghouse 
Plants, NUREG-1431, Rev. 1, dated April 1995. The ten sub-sections 
which the licensee proposes to relocate, without changes to the 
requirements, to the Updated Final Safety Analysis Report or other 
controlled documents relate to: boration system flow path, position 
indication system, rod drop time, seismic instrumentation, chlorine 
detection system, turbine overspeed protection, containment leakage, 
containment structural integrity, electrical equipment protective 
devices and containment penetration conductor overcurrent protective 
devices.
    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 simplify the Technical Specifications (TS), 
meet regulatory requirements for relocated TS, and implement the 
recommendations of the Commission's Final Policy Statement on TS 
Improvements and revised 10 CFR 50.36. Future changes to these 
requirements will be controlled by 10 CFR 50.59. The proposed 
changes are administrative in nature and do not involve any 
modifications to any plant equipment or affect plant operation.
    Therefore, the proposed changes do not involve a significant 
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 are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety-related system performs 
its function. Also, no changes to the operation of the plant or 
equipment are involved.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes involve relocating TS requirements to a 
licensee-controlled document. The requirements to be relocated were 
identified by applying the criteria endorsed in the Commission's 
Final Policy Statement, which is included in the new revision of 10 
CFR 50.36, and are consistent with NUREG-1431, Rev. 1 (Reference 2). 
Thus, the proposed changes do not alter the basic regulatory 
requirements and do not affect any safety analysis.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon

    Date of amendment request: November 2, 1995.
    Description of amendment request: The proposed amendment would 
revise Section 5.0, Administrative Controls, of the Trojan Nuclear 
Plant Technical Specifications, Appendix A to License NPF-1, to reflect 
changes in the organization of the Portland General Electric Company 
(PGE) as they apply to oversite and management of the Trojan Nuclear 
Plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The requested license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The changes in management titles and reporting relationships are 
administrative in nature, do not alter the intent of the Possession 
Only License, and do not modify 

[[Page 58405]]
the present plant systems or adminstrative controls necessary to 
preserve and protect the integrity of the nuclear fuel at the Trojan 
Nuclear Plant. The Trojan Site Executive and Plant General Manager 
will be located at the site and will continue to provide senior 
management attention to each of the functional areas in the Trojan 
Nuclear Plant organization during decommissioning of the facility.
    The general classification of accidents for the permanently 
defueled condition are limited. The three classifications are (1) 
radioactive release from a subsystem or component, (2) fuel handling 
accident, and (3) loss of spent fuel decay heat removal capability. 
The probability of occurrences of consequences from these accidents 
remain unchanged and are bounded by the current accident analysis. 
Therefore, the requested changes do not involve a significant 
increase in the probability or occurrence of an accident previously 
evaluated.
    2. The requested license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The requested amendment is administrative in nature, does not 
affect the manner in which systems and components are operated or 
maintained, and does not alter the intent of the Possession Only 
License. The accident scenarios associated with the permanently 
defueled condition are limited to (1) radioactive release from a 
subsystem or component, (2) fuel handling accident and (3) loss of 
spent fuel decay heat removal capability. There are no new accident 
scenarios or failure modes created by the requested administrative 
changes. Therefore the requested change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The requested license amendment does not involve a 
significant reduction in a margin of safety.
    The requested amendment is administrative in nature, does not 
affect the manner in which systems and components are operated or 
maintained, does not alter the intent of the Possession Only 
License, nor does it adversely impact previously accepted margins of 
safety. Therefore, the requested amendment does not involve a 
significant reduction in margin of safety.

    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensees: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRR Project Director: Seymour H. Weiss.

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

    Date of amendment request: October 7, 1995 as supplemented by 
letter dated October 27, 1995.
    Description of amendment request: The proposed change to Hope Creek 
Technical Specifications (TSs) 4.8.1.1.2, ``A.C. Sources--Operating'', 
would replace the reference to a voltage and frequency band for the 10 
second starting time test with a minimum required voltage and frequency 
that must be attained within 10 seconds.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident [* * *] previously evaluated.
    Since no change is being made to the offsite power supplies, or 
to any system or component that interfaces with the offsite power 
supplies, there is no change in the probability of a Loss of Offsite 
Power Accident.
    Since the proposed change still ensures the surveillance 
requirements meet the licensing basis and since the full spectrum of 
loading, unloading and standby testing performed at the 18 month 
frequency continues to demonstrate the capability of the diesel 
generators to satisfy onsite power requirements during simulated 
accident conditions while the monthly testing demonstrates 
availability, there is no change in the consequences of an accident.
    Since the proposed change will eliminate unnecessary adjustments 
to the governor controls, the probability of malfunction is 
potentially reduced.
    This change ensures the surveillance requirements reflect the 
design basis and provide a basis for consistent timing methodology. 
Since the proposed change is consistent with the intent of the 
existing specifications, and with the design basis of the system and 
since no physical changes are being proposed, no action will occur 
that will increase the probability or consequences of an accident or 
malfunction of equipment important to safety. The diesel generators 
will continue to function as stated in the UFSAR [Updated Final 
Safety Analysis Report].
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident or 
malfunction of equipment important to safety previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change does not result in any design or physical 
configuration changes to the offsite power supplies or to the diesel 
generators. Operation in accordance with the proposed change will 
not impair the diesel generators ability to perform as provided in 
the design basis. By eliminating unnecessary adjustments to the 
diesel generator governor control, performance during any accident 
is potentially enhanced. The diesel generators will continue to 
function as stated in the UFSAR. Therefore, the proposed change will 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    Since the proposed change does not involve the addition or 
modification of plant equipment, is consistent with the intent of 
the existing Technical Specifications, meets the intent of 
applicable Regulatory Guides, and is consistent with the design 
basis of the diesel generators and the UFSAR, no action will occur 
that will 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: M.J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street NW., Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

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: September 29, 1995.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3/4.4.3, Safety Valves and Pilot 
Operated Relief Valve--Operating, and associated Bases 3/4.4.2 and 3/
4.4.3, Safety Valves, to increase the lift setting of the pressurizer 
code safety valves (PSVs) to [equal to or less than] 2575 psig, which 
corresponds to a lift setting tolerance of +3% of the nominal lift 
pressure. Increasing the upper bound of the lift setting tolerance of 
the PSVs from +1% to +3% will allow normal surveillance testing of the 
PSVs to be within +3% of the nominal lift setpoint of 2500 psig, which 
is still acceptable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 

[[Page 58406]]
consideration, which is presented below:

    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 
1 in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because increasing the PSV lift 
tolerance from +1% to +3% only affects the as-found tolerance of the 
PSVs. The initial setting tolerance will still be limited to +1%. No 
hardware modification will be done to the valves which could affect 
any accident initiators.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because increasing the PSV lift 
tolerance from +1% to +3% does not affect the radiological releases 
of any accident previously evaluated in the [Updated Safety Analysis 
Report] USAR. This is not a hardware modification and the reactor 
coolant pressure boundary integrity is unaffected.
    2. Not create the possibility of a new kind of accident from any 
previously evaluated because increasing the PSV lift tolerance from 
+1% to +3% allows the PSVs to protect the reactor coolant pressure 
boundary from overpressure transients. This change only affects the 
allowable lift tolerance. The initial lift setting tolerance is 
still less than +1%. This change does not modify the valve hardware 
or alter the operation of the valves. The possibility of the valves 
spuriously opening during power operation will not be changed. The 
valve setpoint with a -3% lift tolerance is well above the normal 
operating conditions and the [reactor coolant system] RCS high 
pressure trip setpoint.
    3. Not involve a significant reduction in a margin of safety 
because at the +3% lift tolerance the RCS pressure and the reactor 
thermal power are still within the USAR acceptance criteria for a 
control rod withdrawal at low power. This change ensures the 
Technical Specification lift setpoint tolerances are consistent with 
the requirements given in the [American Society of Mechanical 
Engineers] ASME Boiler and Pressure Vessel Code.

    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, William 
Carlson Library, Government Documents Collection, 2801 West 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: Gail H. Marcus.

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

    Date of amendment request: June 21, 1994, as amended by letter 
dated October 23, 1995.
    Description of amendment request: The proposed amendment would 
relocate the review and audit requirements of the On-site Review 
Committee (ORC) and Nuclear Safety Review Board (NSRB) contained in TS 
6.5.1, TS 6.5.2 and TS 6.5.3 to the Operational Quality Assurance 
Manual (OQAM). In addition, the proposed amendment would delete 
reference to the Manager, Nuclear Safety and Emergency Preparedness in 
TS 6.2.3. A revision to the Index was proposed to reflect the 
relocations. This amendment request was previously published in the 
Federal Register on August 31, 1994 (59 FR 45036).
    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 changes are administrative and equivalent descriptions and 
requirements for these oversight committees are contained in the 
OQAM.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    These changes do not involve any physical alterations to the 
plant. There is no new type of accident or malfunction created and 
the method and manner of plant operation will not change. The 
changes are administrative and equivalent descriptions and 
requirements for these oversight committees are contained in the 
OQAM.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety remains unaffected since no design change 
is made and plant operation remains the same. The changes are 
administrative and equivalent descriptions and requirements for 
these oversight committees are contained in the OQAM.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: William H. Bateman.

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: October 17, 1995.
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TS) for the North Anna Power Station, 
Unit No. 2 (NA-2). Specifically, the proposed change would reduce from 
two to one the minimum number of steam generators (SGs) required to be 
opened for inspection during the first refueling outage following an SG 
replacement. TS surveillance requirements 4.4.5.0 through 4.4.5.5 for 
inspection of the SG tubes ensure that the structural integrity of this 
portion of the Reactor Coolant System will be maintained. Accordingly, 
the purpose of TS 4.4.5.1 is to require periodic sample inspections of 
SGs. The initial inspection after SG replacement combined with the 
subsequent inservice inspections serve to provide reasonable assurance 
of detection of structural degradation of the tubes. The proposed TS 
change does not affect or change this basis. However, the requirement 
that two SGs would be opened and inspected during the first refueling 
outage after SG replacement is considered unnecessary.
    The NA-2 SGs were replaced during the first quarter of 1995. The 
purpose of SG replacement was to restore the integrity of the SG tubes 
to a level equivalent to new SGs. In reality, replacement SG components 
incorporate a large number of design improvements which reflect the 
``state-of-the-art'' technology that currently exists for SG design. 
These design improvements will improve the long-term maintainability 
and reliability of the replacement SGs. These enhancements do not 
adversely affect the mechanical or thermal-hydraulic performance of the 
SGs. Thus, the replacement SGs are considered superior to the original 
SGs in terms of design and materials.
    The proposed TS change does not affect or change any limiting 
conditions for operation (LCO) or any other surveillance requirements 
in the TS and the Basis for the surveillance requirement remains 
unchanged. An inspection of the minimum required number of tubes will 
still be performed 

[[Page 58407]]
prior to returning the SGs to service. Although the proposed change 
reduces the number of SGs required to be opened for inspection, the 
minimum number of tubes required to be examined during the inspection 
is not being changed. Thus, the minimum inspected tube population size 
would not be changed.
    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:

    We have evaluated the proposed change against the criteria 
described in 10 CFR 50.92 and concluded that the proposed Technical 
Specifications change does not pose a significant hazards 
consideration.
    [1] The proposed Technical Specifications change does not affect 
the assumptions, design parameters, or results of any UFSAR [Updated 
Final Safety Analysis Report] accident analysis and the proposed 
amendment does not add or modify any existing equipment. Therefore, 
the proposed Technical Specifications change would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    [2] The proposed change to the Technical Specifications does not 
involve modifications to any of the existing equipment or affect the 
operation of any existing systems. The absence of any hardware or 
software changes means that the accident initiators remain 
unaffected, so no unique accident possibility is created. Therefore, 
the proposed Technical Specifications change would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    [3] Although the proposed change will reduce the minimum number 
of steam generators required to be opened for inspection during the 
first refueling outage following steam generator replacement, the 
revised Technical Specification surveillance will continue to ensure 
that a sampling of steam generator tubes will be inspected. The 
operability of the steam generators will also continue to be 
verified by periodic inservice inspections. Therefore, since 
equipment reliability will be maintained, the proposed Technical 
Specifications change will not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews.

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

    Date of amendment request: October 18, 1995.
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) 3.4, ``Steam and Power Conversion System,'' by modifying and 
clarifying the operability requirements for the main steam safety 
valves (MSSVs), the auxiliary feedwater (AFW) System, and the 
condensate storage tank system.
    The proposed amendment would eliminate inconsistencies within TS 
Section 3.4 and provide the basis for acceptable operation of the 
Auxiliary Feedwater System below 15% reactor power. The proposed 
amendment supersedes in its entirety a previously submitted proposed 
amendment dated May 20, 1994, which was noticed in the Federal Register 
on September 28, 1994 (59 FR 49442).
    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:

Significant Hazards Determination for Proposed Changes to Technical 
Specification (TS) 3.4.a ``Main Steam Safety Valves''

    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Currently, TS 3.4.a.1.A.2 requires five MSSVs to be operable 
prior to heating the reactor > 350  deg.F. The proposed change 
requires a minimum of two MSSVs per steam generator to be operable 
prior to heating the reactor coolant system > 350  deg.F, and five 
MSSVs per steam generator to be operable prior to reactor 
criticality. If these conditions cannot be met within 48 hours, 
within 1 hour action shall be initiated to achieve hot standby 
within 6 hours, achieve hot shutdown within the following 6 hours, 
and achieve and maintain the reactor coolant system temperature < 
350  deg.F within an additional 12 hours.
    The MSSVs are relied upon to function in each of the following 
USAR analyzed accidents: Reactor Coolant Pump Locked Rotor, Loss of 
External Electrical Load, Loss of Normal Feedwater, Uncontrolled Rod 
Cluster Control Assembly Withdrawal, Steam Generator Tube Rupture, 
and Anticipated Transients without Scram.
    In a subcritical condition, two operable MSSVs are capable of 
relieving the maximum steam generated during these anticipated 
design basis transient events. Because this proposed TS requires all 
MSSVs to be operable prior to reactor criticality, there will be no 
adverse effect on the health and safety of the public.
    In all cases, the relieving capacity of the MSSVs is sufficient 
to maintain steam pressures within safety analysis acceptable 
criteria, and reactor criticality is not permitted unless all MSSVs 
are operable. Therefore, there is no adverse effect on the health 
and safety of the public and no significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
operating setpoints, or overall plant performance. Therefore, it 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The USAR safety analysis assumes five MSSVs per steam generator 
are operable. However, as shown above, this change results in no 
steam generator overpressure event or increase in the radiological 
dose. Therefore, this change will not involve a reduction in the 
margin of safety.

Significant Hazards Determination for Proposed Changes to Technical 
Specification (TS) 3.4.b ``Auxiliary Feedwater System''

    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Current TS 3.4.a.1.A.1 and TS 3.4.b governing auxiliary 
feedwater flow to the steam generators are being combined and 
titled, ``Auxiliary Feedwater System.'' This change is consistent 
with the format of ``Westinghouse Standard Technical 
Specifications,'' NUREG-1431. In addition to the formatting changes, 
a number of technical changes are being proposed. These are:
    The correction of an inconsistency between current TS 
3.4.a.1.A.1 and current TS 3.4.b.2.A.
    The addition of a seven (7) day Limiting Condition for Operation 
(LCO) action statement for one inoperable steam supply to the 
turbine driven auxiliary feedwater pump.
    A specification is being added to permit any of the following 
conditions with reactor power less than 15%, without declaring the 
corresponding AFW train inoperable: the AFW pump control switches 
located in the control room to be in the ``pullout'' position, flow 
control valves AFW-2A and AFW-2B to be in a throttled or closed 
position, and train cross-connect valves AFW-10A and AFW-10B to be 
in the closed position.
    An inconsistency currently exists between current TS 3.4.a.1.A.1 
and current TS 

[[Page 58408]]
3.4.b.2.A. TS 3.4.a.1.A.1 requires the system piping and valves 
directly associated with providing auxiliary feedwater flow to the 
steam generators to be operable, with a corresponding 48 hour 
limiting condition for operation (LCO) action statement if this 
requirement is not met. TS 3.4.b.2.A allows one auxiliary feedwater 
pump to be inoperable for 72 hours. This arrangement can cause a 
conflict regarding which TS is applicable depending on which 
component in the auxiliary feedwater flowpath to the steam 
generators is inoperable. By moving all TS action statements to TS 
3.4.b, the inconsistency between TS 3.4.a.1.A.1 and TS 3.4.b.2.A 
will be eliminated. The requirement to maintain the operability of 
the system piping and valves directly associated with providing 
auxiliary feedwater flow to the steam generators remains, but is 
being modified to prevent the removal of both AFW supply headers 
from service.
    Proposed TS 3.4.b.2.C is being added to allow one steam supply 
to the turbine driven auxiliary feedwater pump to be inoperable for 
seven days. This addition is consistent with ``Westinghouse Standard 
Technical Specifications,'' NUREG-1431. The seven day completion 
time is reasonable based on the redundant steam supplies to the 
pump, the availability of the redundant motor-driven AFW pumps, and 
the low probability of an event occurring that requires the 
inoperable steam supply to the turbine driven AFW pump. For these 
reasons, this change will have no adverse effect on the health and 
safety of the public.
    Proposed TS 3.4.b.6.A and B permit the AFW Pump control switches 
located in the control room to be placed in the ``pull out'' 
position and valves AFW-2A and AFW-2B to be in a throttled position 
when below 15% reactor power without declaring the corresponding AFW 
train inoperable. This change is proposed to resolve concerns 
regarding the cycling of the AFW pumps and the throttling of valves 
AFW-2A and AFW-2B during plant startups and shutdowns. Analysis 
shows that control room operators have a minimum of ten minutes to 
initiate auxiliary feedwater flow after a design basis accident with 
no steam generator dryout or core damage.
    All accidents which rely on AFW flow for mitigation were 
reanalyzed to support this change. These analyses were completed 
assuming an initial power of 100%. However, a 15% reactor power 
restriction has been imposed on placing the AFW pump control 
switches located in the control room in the ``pull out'' position 
and throttling valves AFW-2A and AFW-2B. This restriction in effect 
limits use of TS 3.4.b.6 to plant startups, shutdowns and other low 
power operating conditions.
    This change alters the assumptions of the safety analysis for 
the Small-Break Loss of Coolant Accident, the Steam Generator Tube 
Rupture and the Loss of Normal Feedwater due to their dependence on 
the AFW system to start and supply AFW for heat removal. To support 
this change, the Westinghouse Electric Corporation performed an 
analysis of the Small-Break Loss-of-Coolant Accident using the 
NOTRUMP code assuming ten minutes for operator action to initiate 
auxiliary feedwater. This analysis resulted in a Peak Cladding 
Temperature (PCT) of 1053  deg.F from an initial power level of 
100%. In addition, all other acceptance criteria of 10 CFR 50.46 
were met. This large margin to the 2200  deg.F PCT limit supports 
ten minutes for operator action to initiate auxiliary feedwater.
    Furthermore, WPSC has analyzed the Loss of Normal Feedwater and 
the Steam Generator Tube Rupture Accident assuming delays in the 
initiation of auxiliary feedwater. The Loss of Normal Feedwater 
Accident with a ten minute delay in the initiation of Auxiliary 
Feedwater does not result in any adverse condition in the core. It 
does not result in water relief from the pressurizer safety valves, 
nor does it result in uncovering the tube sheets of the steam 
generators. Also, at all times the Departure from Nucleate Boiling 
Ratio (DNBR) remained greater than 1.30. The Steam Generator Tube 
Rupture Accident with no auxiliary feedwater flow was also analyzed. 
The results of this analysis indicate that neither steam generator 
empties of liquid and at least 20  deg.F of reactor coolant system 
subcooling is maintained throughout the transient. Also, there is no 
increase in the radiological dose to the public.
    Ten minutes is an acceptable time for operator action because 
four independent alarms in the control room would initiate operator 
action to place the AFW pump control switches to the ``auto'' 
position and initiate AFW flow to the steam generators when 
necessary. These include two steam generator lo level alarms (one 
per steam generator), and two steam generator lo-lo level alarms 
(one per steam generator). Provisions also exist to add additional 
low level alarms on the plant process computer. In addition to these 
alarms, control room operators have twelve other indications of 
insufficient, or no, AFW flow to the steam generators. These 
indications include three auxiliary feedwater pump low discharge 
pressure alarms (one per AFW pump), two auxiliary feedwater flow 
meters (one per steam generator), two AFW pump motor amp meters (one 
per motor-driven AFW pump), two ``ESF in Pullout'' alarms (one per 
Engineered Safety Features train) and three pump running lights (one 
per AFW pump). The ten minutes for operator action was discussed in 
a telephone conversation between WPSC and Mr. R. Laufer (NRR). Ten 
minutes for operator action is further supported by Branch Technical 
Position EISCB 18. Scenarios have been completed on the KNPP 
simulator to support ten minutes for operator initiation of AFW 
flow. In all cases, operators manually initiated AFW flow within the 
allowed ten minutes.
    Proposed TS 3.4.b.6.C permits valves AFW-10A and AFW-10B to be 
in the closed position when below 15% reactor power without 
declaring the turbine-driven AFW train inoperable. This change is 
being proposed to allow operational flexibility of the AFW system 
during startups and shutdowns. As described below, the operability 
of the turbine-driven auxiliary feedwater train is independent of 
the position of the valves AFW-10A and AFW-10B. However, the 
operability of this train is dependent on the ability of these 
valves to reposition.
    The operability of the AFW system following a main steam line 
break (MSLB) was reviewed in our response to IE Bulletin 80-04. As a 
result of this review, requirements for the turbine-driven AFW pump 
were originally added to the Technical Specifications.
    For all other design basis accidents, the two motor-driven AFW 
pumps supply sufficient redundancy to meet single failure criteria. 
In a secondary line break, it is assumed that the pump discharging 
to the intact steam generator fails and that the flow from the 
redundant motor-driven AFW pump is discharging out the break. 
Therefore, to meet single failure criteria the turbine-driven AFW 
pump was added to Technical Specifications.
    The cross-connect valves (AFW-10A and AFW-10B) are normally 
maintained in the open position. This provides an added degree of 
redundancy above what is required for all accidents except for a 
MSLB. During a MSLB, one of the cross-connect valves will have to be 
repositioned regardless if the valves are normally open or closed. 
Therefore, the position of the cross-connect valves does not affect 
the operability of the turbine-driven AFW train. However, 
operability of the train is dependent on the ability of the valves 
to reposition.
    For these reasons, this change will have no adverse effect on 
the health and safety of the public or significantly increase the 
probability or consequences of an accident previously evaluated in 
the USAR.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The auxiliary feedwater system is required to mitigate the 
consequences of an accident. The auxiliary feedwater system is not 
an accident initiator. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    This change alters the assumptions of the safety analysis for 
the Small-Break Loss-of-Coolant Accident, the Steam Generator Tube 
Rupture and the Loss of Normal Feedwater due to their dependence on 
the AFW system to start and supply AFW flow for heat removal. To 
support this change the Westinghouse Electric Corporation has 
performed an analysis of the Small-Break Loss-of-Coolant Accident 
using the NOTRUMP code assuming ten minutes for operator action to 
initiate auxiliary feedwater. This analysis resulted in a Peak 
Cladding Temperature (PCT) of 1053 deg. F from an initial power 
level of 100%. In addition, all other acceptance criteria of 10 CFR 
50.46 were met. This large margin to the 2200 deg. F PCT limit 
supports ten minutes for operator action to initiate auxiliary 
feedwater.
    Furthermore, WPSC has analyzed the Loss of Normal Feedwater and 
the Steam Generator Tube Rupture Accident assuming delays in the 
initiation of auxiliary feedwater. The Loss of Normal Feedwater 
Accident with a ten-minute delay in the initiation of Auxiliary 
Feedwater does not result in any adverse condition in the core. 

[[Page 58409]]
It does not result in water relief from the pressurizer safety valves, 
nor does it result in uncovering the tube sheets of the steam 
generators. Also, at all times the Departure from Nucleate Boiling 
Ratio (DNBR) remained greater than 1.30. The Steam Generator Tube 
Rupture Accident with no Auxiliary Feedwater flow was also analyzed. 
The results of this analysis indicate that neither steam generator 
empties of liquid and at least 20 deg. F of reactor coolant system 
subcooling is maintained throughout the transient. Also, there is no 
increase in the radiological dose to the public. For these reasons, 
these changes will not adversely affect the health and safety of the 
public or involve a significant reduction in the margin of safety.
    As discussed in the safety evaluation, the operability of the 
turbine-driven AFW train is independent of the position of valves 
AFW-10A and AFW-10B. However, the operability of the train is 
dependent on the ability of these valves to be repositioned. 
Therefore, the proposed change has no impact on the accident 
analysis and no effect on the margin of safety.

Significant Hazards Determination for Proposed Administrative 
Changes to Section TS 3.4, ``Steam and Power Conversion System''

    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will 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 the margin of safety.
    The proposed changes are administrative in nature and do not 
alter the intent or interpretation of the TS. Therefore, no 
significant hazards exist.
    Additionally, the proposed change is similar to example C.2.e(i) 
in 51 FR 7751. Example C.2.e.(i) states that changes which are 
purely administrative in nature; i.e., to achieve consistency 
throughout the Technical Specifications, correct an error, or a 
change in nomenclature, are not likely to involve a significant 
hazard.

    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 Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
PO Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: October 18, 1995.
    Description of amendment request: This license amendment would 
replace the current fuel oil volume requirement in the emergency diesel 
generator (EDG) day tank in Technical Specifications 3.8.1.1.b.1) and 
3.8.1.2.b.1) with a fuel oil level requirement. Associated Surveillance 
Requirement 4.8.1.1.2.a.1) would also be changed to replace the 
requirement to visually check the fuel oil level in the day tank with a 
requirement to verify that the fuel oil transfer pump starts on low 
level in the day tank standpipe.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will increase the minimum amount of diesel 
fuel oil that the current specifications require to be maintained in 
the EDG day tanks for standby operation. This change reflects the 
level that has been administratively maintained since the beginning 
of plant operation. The proposed change will not affect the way the 
EDG is operated and does not affect the ability of the EDGs to 
perform their safety function. The surveillance requirement change 
is being made to more thoroughly reflect the method used to assure 
the tank level is being properly maintained. The proposed change 
will not require the EDG to be operated in a manner different than 
that for which it was designed. Therefore, the proposed change will 
not significantly increase the consequences of an accident or 
malfunction of equipment important to safety previously evaluated in 
the USAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no active components being added whose failure could 
prevent the EDG from functioning. There is no new type of accident 
or malfunction being created and the method and manner of plant 
operation remains unchanged. The safety design bases in the USAR 
have not been altered. Thus, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    No new or different accident scenarios, transient precursors, 
failure mechanisms, or limiting single failures will be introduced 
as a result of these changes. The method of operation of the EDGs is 
not being altered, and the fuel oil transfer pumps will continue to 
perform the same function they currently perform. Therefore, the 
possibility of a new or different kind of accident other than those 
already evaluated will not be created by this change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There are no changes being made to any safety limits or safety 
system settings that would adversely impact plant safety. Although 
the minimum required amount of fuel oil specified in the Technical 
Specifications is being revised, this amount of fuel oil has been 
administratively controlled since the beginning of commercial 
operation. Thus, the operability of the emergency diesel generators 
has never been affected by this issue. Neither the method of 
operation of the EDGs nor their safety function are being altered by 
the proposed change. Therefore, the proposed change would not result 
in 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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: October 24, 1995.
    Description of amendment request: This license amendment request 
proposes to revise Surveillance Requirement 4.7.6.e.4 to reflect a 
design change, scheduled to be installed during the next refueling 
outage, that would change the output rating of the charcoal filter 
adsorber unit heater in the pressurization portion of the control room 
emergency ventilation system (CREVS) from 15 kW to 5 kW. Proposed 
revisions to Surveillance Requirements 4.7.6.c.2 and 4.7.6.d are 
included which would change the acceptance criteria for the testing of 
carbon samples from the CREVS charcoal adsorbers. The proposal would 
adapt ASTM D 3803-1989 as the laboratory testing standard with the 
testing to be performed at 30 degrees Centigrade and 70 percent 

[[Page 58410]]
relative humidity for a methyl iodide penetration of 2 percent.
    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 design function of the filter adsorber unit heater in the 
pressurization system portion of CREVS is to reduce the relative 
humidity of the air entering the charcoal filter beds to 70% 
relative humidity. Although the original design specified a heater 
with a rating of 15 kW, review of the design basis calculation for 
this system indicates that only 2.09 kW is actually required 
(including applicable margins to allow for voltage variations). The 
proposed change to the CREVS heaters' output rating from 15 kW to 5 
kW will not affect the method of operation of the system, and the 
new heater capacity will still exceed filter operational 
requirements and safety margin. Neither the heater change nor the 
charcoal testing protocol changes will affect system operation or 
performance, nor do they affect the probability of any event 
initiators. These changes do not affect any Engineered Safety 
Features actuation setpoints or accident mitigation capabilities. 
Therefore, the proposed changes will not significantly increase the 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated in the USAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested change to the CREVS heaters' output rating and the 
changes to the charcoal sample testing protocol will not affect the 
method of operation of the system, and the new heater capacity will 
still exceed filter operational requirements and safety margin by a 
significant amount. The proposed changes only affect the heater size 
in the system and the testing criteria for the charcoal samples. No 
new or different accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of these changes. Therefore, the possibility of a new or 
different kind of accident other than those already evaluated will 
not be created by this change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The requested change to the CREVS heaters' output rating will 
reduce the heater output of the system, but the new heater capacity 
will still exceed filter operational requirements and safety margin 
by a significant amount. In addition, the reduction in heat load 
output from the heater will increase the design margin between the 
cooling capacity of the system air conditioning units and the 
building heat load. The new charcoal adsorber sample laboratory 
testing protocol is more stringent than the current testing practice 
and more accurately demonstrates the required performance of the 
adsorbers following a design basis LOCA [loss-of-coolant accident]. 
Therefore, these changes will not reduce the margin of safety of the 
CREVS filter operation.

    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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.

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

    Date of amendment request: September 20, 1995.
    Description of amendment request: The proposed amendment would 
modify the Appendix A Technical Specifications for the Engineered 
Safety Features Actuation System (ESFAS) Instrumentation. Specifically, 
the proposed amendment would revise the Seabrook Station Technical 
Specifications to relocate Functional Unit 6.b, ``Feedwater Isolation--
Low RCS Tavg Coincident with a Reactor Trip'' from Technical 
Specification 3.3.2. ``Engineered Safety Features Actuation System 
Instrumentation'' to the Seabrook Station Technical Requirements Manual 
which is a licensee controlled document.
    Date of publication of individual notice in Federal Register: 
October 24, 1995 (60 FR 54524).
    Expiration date of individual notice: November 24, 1995.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 58411]]


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

    Date of application for amendments: October 25, 1994, as 
supplemented by letter dated September 11, 1995.
    Brief Description of amendments: The proposed amendments change the 
Technical Specifications to relocate the remaining Environmental 
Technical Specifications to other licensee-controlled documents and 
delete the 30-day reporting requirement for inoperable meteorological 
instrumentation.
    Date of issuance: November 2, 1995.
    Effective date: November 2, 1995.
    Amendment Nos.: 179 and 210.
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63113). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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: May 13, 1993 as supplemented 
August 11 and September 20, 1995.
    Brief description of amendments: The amendments revised Section 3/
4.6.1.7 of the Technical Specifications, Containment Purge Ventilation 
System, to allow the simultaneous opening of the 8-inch miniflow purge 
supply and exhaust valves to ensure the containment atmosphere is 
conducive to human occupants and to maintain their dose as low as 
reasonably achievable.
    Date of issuance: November 2, 1995.
    Effective date: November 2, 1995.
    Amendment Nos.: 76, 76, 68, and 68.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48379). The August 11 and September 20, 1995, submittals provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 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: September 1, 1995, as 
supplemented on September 1 (two letters), September 2, September 4, 
September 8, September 15, September 19, September 20, September 22, 
October 3, October 7, October 11 (two letters), October 13 (three 
letters), October 23 and October 26, 1995.
    Brief description of amendments: The amendments revise the steam 
generator (SG) repair criteria in the Byron, Unit 1 and Braidwood, Unit 
1 Technical Specifications. These revisions add a set of voltage-based 
SG tube repair criteria different from those previously added by 
License Amendment No. 66, dated October 24, 1994, to the Byron 1 TSs 
and by License Amendment No. 54, dated August 18, 1994, to the 
Braidwood 1 TSs. The present set of voltage repair limits which are 
being added to the Byron 1 and Braidwood 1 TSs are applicable only for 
a specific form of SG tube degradation identified as outer diameter 
stress corrosion cracking (ODSCC) which is confined entirely within the 
thickness of the tube support plates (TSPs) in the SGs. The voltage-
based repair criteria for the cold-leg side of the SGs for SG tubes 
with ODSCC indications and for SG tubes on the hot-leg side which show 
significant denting, are consistent with those provided in the NRC 
staff's guidance contained in Generic Letter 95-05, dated August 3, 
1994.
    The lower voltage repair limit for the SG tubes with ODSCC 
indications on the hot-leg side of the SGs have been raised from 1.0 to 
3.0 volts as measured by a bobbin coil. All bobbin indications below 
3.0 volts will be allowed to remain in service and all bobbin 
indications above this limit will be either repaired or removed from 
service by plugging.
    This revision to the voltage repair limits on the hot-leg side 
reflects a methodology which is significantly different than that 
contained in GL 95-05. The principal difference between the methodology 
being applied for the 3.0 volt criteria on the hot-leg side is that the 
Commonwealth Edison Company (ComEd) is taking credit for the constraint 
provided by the TSPs to reduce the probability of SG tube burst in the 
event of a severe accident (i.e., a main steamline break). This 
constraint is assured by modifying a limited number of SG tubes so that 
they provide additional stiffness to the TSPs, thereby reducing to a 
small amount, their deflection under MSLB blowdown loads.
    Additionally, inspection and reporting requirements are being added 
to the Byron 1 and Braidwood 1 TSs in support of the revised voltage-
based repair criteria. Further, the maximum permissible value of the 
iodine-131 concentration in the primary coolant in the Byron 1 TSs is 
reduced from 1.0 to 0.35 microcuries per gram of coolant. This is the 
same value for the iodine-131 primary coolant concentration in the 
Braidwood 1 TSs. Finally, the Bases sections in the Byron 1 and 
Braidwood 1 TSs are revised to provide a concise description of the 
methodology proposed by ComEd in support of its proposed revision of 
the voltage-based SG tube repair criteria.
    Date of issuance: November 9, 1995.
    Effective date: November 9, 1995.
    Amendment Nos.: 77, 77, 69, and 69.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49963).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 9, 1995. The supplemental 
submittals listed above provide clarifying technical information that 
does not affect the initial No Significant Hazards Consideration 
Determination.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481. 

[[Page 58412]]


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

    Date of application for amendment: July 5, 1995.
    Brief description of amendment: This amendment revises Section 6.0 
of the Technical Specifications to incorporate several administrative 
controls and editorial changes to the Training, Plant Review Committee, 
and Plant Safety and Licensing staff sections.
    Date of issuance: November 3, 1995.
    Effective date: November 3, 1995.
    Amendment No.: 170.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39435).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendments: April 10, 1995.
    Brief description of amendments: The amendments revise the required 
number of operable hydrogen igniters to allow removal of two hydrogen 
igniters serving the lower reactor cavity and incore instrument cable 
tunnel.
    Date of issuance: October 30, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 136 and 130.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49932).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments: September 13, 1995.
    Brief description of amendments: The amendments modify the notation 
for the overpower delta temperature reactor trip heatup setpoint 
penalty coefficient as delineated in Note 3 in Technical Specification 
Table 2.2-1 in order to make the nomenclature consistent with the 
Standard Technical Specifications and to facilitate a modification to 
reduce the reactor coolant system hot leg temperature as planned during 
the Catawba Unit 2 end-of-cycle 7 refueling outage.
    Date of issuance: October 31, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: 137 and 131.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49933).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments: September 1, 1995, as 
supplemented October 17, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 6.9.1.9 to include references to updated or recently 
approved methodologies used to calculate cycle-specific limits 
contained in the Core Operating Limits Report. The subject references 
have been reviewed and approved by the NRC staff.
    Date of issuance: November 2, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: 138 and 132.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49932). The October 17, 1995, letter provided clarifying information 
that did not change the scope of the September 1, 1995 application and 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application for amendments: June 13, 1994, as supplemented 
by letters dated August 15, 1994, March 23, April 18, July 21, and 
September 22, 1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase the initial fuel enrichment limit 
and establish new loading patterns for new and irradiated fuel in the 
spent fuel pool to accommodate this increase.
    The March 23, 1995, supplement, which provided additional 
information that modified the June 13, 1994, application's no 
significant hazards consideration determination, also revises the TS to 
(1) change the surveillance requirement for boron concentration in the 
spent fuel pool (SFP), (2) remove the option to use alternate storage 
configurations in the SFP and replace it with footnotes, (3) add 
information contained in the Bases to the footnotes, and (4) change the 
Bases to discuss the option to use specific analyses on alternate fuel.
    Date of issuance: November 6, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 159 and 141.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8746); and May 8, 1995 (60 FR 22590). The April 18, July 21, and 
September 22, 1995, letters provided additional clarifying information 
that did not change the scope of the June 13, 1994, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 6, 1995, and Environmental 
Assessment dated August 17, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223. 

[[Page 58413]]


Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: May 17, 1995.
    Brief description of amendment: The amendment will extend the 
applicability of the current Reactor Coolant System (RCS) Pressure/
Temperature Limits and maximum allowed RCS heatup and cooldown rates to 
23.6 Effective Full Power Years (EFPY) of operation. In addition, 
administrative changes were proposed for TS 3.1.2.1 (Boration Systems 
Flow Paths-Shutdown) and TS 3.1.2.3 (Charging Pump-Shutdown) to clarify 
the conditions for which a High Pressure Safety Injection pump may be 
used.
    Date of Issuance: October 27, 1995.
    Effective Date: October 27, 1995.
    Amendment No.: 141.
    Facility Operating License No. DPR-67: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32362).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 27, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendments: February 28, 1994.
    Brief description of amendments: The amendments delete the minimum 
frequency criteria prescribed for quality assurance audits from 
Administrative Controls sections 6.5.2.8 and 6.8.4 of the Technical 
Specifications (TS). Audit periodicity will thereby be controlled by 
the program described in the Florida Power and Light Company (FPL) 
Topical Quality Assurance Report.
    Date of Issuance: October 25, 1995.
    Effective Date: October 25, 1995.
    Amendment Nos.: 140 and 80.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17599).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 25, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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: July 26, 1995.
    Brief description of amendments: These amendments revise selected 
line items from NRC Generic Letter 93-05, ``Line-Item Technical 
Specification Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation.''
    Date of issuance: October 17, 1995.
    Effective date: October 17, 1995.
    Amendment Nos.: 177 and 171.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47617).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 17, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of amendment request: December 10, 1993.
    Brief description of amendment: The amendment revises the Cooper 
Nuclear Station Technical Specifications to change the reporting 
frequency of the Radioactive Materials Release Report from semiannual 
to annual and to extend the reporting frequency of the Annual Design 
Change Report from annual to annually or along with the Updated Safety 
Analysis Report updates required by 10 CFR 50.71(e). This change 
reflects revised requirements contained in 10 CFR 50.36a and 10 CFR 
50.59(b).
    Date of issuance: November 3, 1995.
    Effective date: November 3, 1995.
    Amendment No.: 172.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7691).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated Novemver 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305.

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

    Date of amendment request: June 28, 1995.
    Brief description of amendment: The amendment revises the Cooper 
Nuclear Station Technical Specifications to increase the required 
reactor pressure vessel boron concentration, to modify the surveillance 
frequency for standby liquid control system pump operability testing 
from monthly to quarterly, and to make editorial changes.
    Date of issuance: November 8, 1995.
    Effective date: November 8, 1995.
    Amendment No.: 173.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39441).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 8, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305.

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

    Date of amendment request: September 5, 1995.
    Description of amendment request: The amendment modifies the 
Appendix A Technical Specifications (TSs) for the Turbine Cycle Safety 
Valves. Specifically, the amendment changes Seabrook Station Appendix A 
Technical Specification Table 3.7-1 to reduce the Maximum Allowable 
Power Range Neutron Flux--High Setpoints with Inoperable Main Steam 
Safety Valves (MSSVs) and Table 3.7-2 to reduce the opening setpoints 
of the MSSVs. Bases Section 3/4.7.1.1 is changed to include the 
algorithm used for determining the new setpoint values.
    Date of issuance: November 2, 1995.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 43.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 2, 1995 (60 FR 
51505). 

[[Page 58414]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, New Hampshire 03833.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: December 21, 1994, as 
supplemented February 22, 1995.
    Brief description of amendment: The amendment revises the License 
Condition C.(3), Fire Protection, and certain of the Technical 
Specifications (TS) related to fire protection requirements. The 
amendment changes the TS by relocating them to another controlled 
document, the Technical Requirements Manual referenced in the Final 
Safety Analysis Report.
    Date of issuance: November 3, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 191.
    Facility Operating License No. DPR-65: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6303) The February 22, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: August 31, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to remove the phrase ``other than Millstone Unit No. 2'' 
from the Administrative Controls Section 6.3.1, Item (a). This relates 
to Amendment No. 178 that changed the Technical Specifications to 
require an individual who serves as the Operations Manager to either 
hold a Millstone Unit 2 Senior Reactor Operator (SRO) license or have 
held an SRO license at another pressurized water reactor other than the 
Millstone Unit No. 2. If the Operations Manager does not hold a 
Millstone Unit No. 2 SRO license, then an individual serving as the 
Assistant Operations Manager would be required to possess an SRO 
license at Millstone Unit 2.
    Date of issuance: November 2, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 190.
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49941).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of application for amendments: January 27, 1995.
    Brief description of amendments: The amendments change the Limerick 
Generating Station Units 1 and 2 Technical Specifications (TS) by 
eliminating the TS active safety function designation of eight (i.e., 
four per unit) Drywell Chilled Water System valves.
    Date of issuance: October 30, 1995.
    Effective date: October 30, 1995.
    Amendment Nos.: 103 and 67.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20524).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: November 23, 1994, as 
supplemented by letter dated August 31, 1995.
    Brief description of amendment: The proposed changes to the 
Technical Specifications (TSs) revise TS 4.8.2.1, ``Electrical Power 
Systems--D.C. Sources,'' Surveillance Requirements, and associated 
Bases Section 3/4.8.2.
    Date of issuance: October 31, 1995.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 87.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39449). The August 31, 1995, letter provided additional and clarifying 
information that did not change the scope of the November 23, 1994, 
application and the initial proposed no significant consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.
    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: November 28, 1994.
    Brief description of amendment: This amendment revises the 
technical specifications for the Reactor Coolant System recirculation 
flow upscale trip function to change the trip setpoint and allowable 
value to reflect 105% of rated core flow, item one of the above 
application.
    Date of issuance: October 31, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 86.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications. 

[[Page 58415]]

    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39450).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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

    Date of application for amendments: March 30, 1995, as supplemented 
August 18, 1995.
    Brief description of amendments: The amendments eliminate the 
defined term CONTROLLED LEAKAGE, remove Controlled Leakage flow from 
the Reactor Coolant System Operational Leakage Limiting Condition for 
Operation (LCO) and establish a new Seal Injection Flow LCO.
    Date of issuance: October 30, 1995.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 178 and 159.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 1995 (60 FR 
24918). The August 18, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendments: August 1, 1995, as supplemented 
by letter dated October 18, 1995.
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3/4.3.2, ``Engineered Safety Features Actuation 
System Instrumentation,'' Table 3.3-3. Table 3.3-3 includes the 
requirements for the minimum number of toxic gas isolation system 
(TGIS) trains operable. These amendments are a one-time-only change to 
extend the allowed TGIS outage times during the replacement of the 
existing TGIS instrumentation.
    Date of issuance: November 2, 1995.
    Effective date: November 2, 1995, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--Amendment No. 126; Unit 3--Amendment No. 
115.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47625). The October 18, 1995, supplemental letter provided 
clarifying information and did not change the initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

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

    Date of application for amendments: September 30, 1993 (TS-337).
    Brief Description of amendment: The amendments revise the operating 
license to reflect issuance of a safety evaluation dated November 2, 
1995 accepting the revised Appendix R Safe Shutdown Program to 
accommodate simultaneous power operation of Browns Ferry Units 2 and 3.
    Date of issuance: November 2, 1995.
    Effective Date: November 2, 1995.
    Amendment Nos.: 226, 241 and 200.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
629).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room Location: Athens Public library, South 
Street, Athens, Alabama 35611.

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

    Date of application for amendments: January 4, 1995 (TS 355).
    Brief Description of amendment: The amendments revise applicability 
and surveillance requirements for the intermediate power range monitor, 
average power range monitor (APRM), and APRM Inoperative Trip 
functions.
    Date of issuance: November 2, 1995.
    Effective Date: November 2, 1995.
    Amendment Nos.: 227, 242 and 201.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29888).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room Location: Athens Public library, South 
Street, Athens, Alabama 35611.

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

    Date of application for amendments: June 2, 1995 (TS 361/371).
    Brief Description of amendment: The amendments revise the 
operability definition for residual heat removal service water 
components for use as a standby coolant supply. The amendments also 
incorporate related changes to the technical specification Bases which 
were submitted on October 2, 1995.
    Date of issuance: November 2, 1995.
    Effective Date: November 2, 1995.
    Amendment Nos.: 225, 240 and 199.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42610).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 2, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room Location: Athens Public library, South 
Street, Athens, Alabama 35611.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendment: May 19, 1995; revised September 
11, 1995 (TS 95-13). 

[[Page 58416]]

    Brief description of amendment: The amendment modifies License 
Condition 2.C.(17) by extending the required surveillance interval to 
May 18, 1996, for Surveillance Requirement 4.3.2.1.3 for certain 
specified engineered safety features response time tests.
    Date of issuance: October 30, 1995.
    Effective date: October 30, 1995.
    Amendment No.: 204.
    Facility Operating License No. DPR-79: Amendment revises the 
operating license.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32372); renoticed September 27, 1995 (60 FR 49948).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 30, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

    Dated at Rockville, Maryland, this 15th day of November 1995.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 95-28606 Filed 11-24-95; 8:45 am]
BILLING CODE 7590-01-P