[Federal Register Volume 59, Number 187 (Wednesday, September 28, 1994)]
[Unknown Section]
[Page ]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-23819]


[Federal Register: September 28, 1994]


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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 September 2, 1994, through September 16, 
1994. The last biweekly notice was published on September 14, 1994 (59 
FR 47163).

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 
20555. The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By October 28, 1994, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why internvetion should be permitted with particular reference to the 
following factors: (1) The Nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(l)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendment request: August 25, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 2.2-1, Reactor Trip System 
Instrumentation Trip Setpoints, and Table 3.3-4, Engineered Safety 
Actuation System Instrumentation Trip Setpoints, to reflect a revised 
steam generator [water] level process measurement accuracy. The steam 
generator level trip setpoints are not affected by the proposed 
amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The Technical Specification Tables 2.2-1 and 3.3-4 which 
document Total Allowable, Z, and S values are being revised to 
reflect additional Process Measurement Error uncertainties based 
upon enhanced knowledge of steam generator performance provided by 
the Nuclear Steam Supply System (NSSS) vendor. There will be no 
physical changes to plant equipment, logic, or control as a result 
of the proposed amendment. The safety-related trip setpoints are not 
being changed. Therefore, there would be no increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The propose amendment does not introduce any new equipment, 
logic, or control functions. The steam generator water level 
protective setpoints are not being changed. No new common mode 
failure mechanism is being introduced. 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 amendment does not involve a significant 
reduction in the margin of safety.
    The proposed amendment revises Technical Specification Tables 
2.2-1 and 3.3-4 to more accurately reflect PMA uncertainties based 
upon enhanced knowledge of the steam generator performance provided 
by the NSSS vendor. The margin of safety as defined in the Technical 
Specifications is not reduced by the proposed changes to the Tables. 
Calculations demonstrate that this requirement is still satisfied 
with the new values based upon the enhanced understanding of PMA 
terms. Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: August 25, 1994.
    Description of amendment request: The requested amendments allow 
the testing interval for auxiliary feedwater (AFW) system pumps to be 
increased from monthly to quarterly on a staggered test basis. The 
proposed amendments are consistent with NRC staff recommendations and 
guidance contained in NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements'' and Generic Letter 93-05, 
``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation.''
    In addition, the requested amendments incorporate a note from 
Surveillance Requirement 3.7.5.2 of NUREG-1431, ``Revised Standard 
Technical Specifications, Westinghouse Plants'' into the existing 
McGuire Technical Specifications governing AFW system pump testing. 
This note clarifies that the turbine-driven AFW pump cannot be tested 
until the required pressure exists in the secondary side of the steam 
generator.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    As required by 10 CFR 50.91, this analysis is provided 
concerning whether the requested amendments involve significant 
hazards considerations, as defined by 10 CFR 50.92. Standards for 
determination that an amendment request involves no significant 
hazards considerations are if operation of the facility in 
accordance with the requested 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 requested amendments decrease from monthly to quarterly the 
frequency at which the motor-driven and turbine-driven AFW pumps 
must be demonstrated operable as specified in TS... 4.7.1.2 
(McGuire). They also incorporate a note of clarification from the 
new Westinghouse STS [Standard Technical Specifications] into the 
existing... McGuire specification concerning when the pump head or 
discharge pressure versus flow verification for the turbine-driven 
pump is required to be performed.
    In 48 FR 14870, the Commission has set forth examples of 
amendments that are considered not likely to involve significant 
hazards considerations. Example vii describes a change to make a 
license conform to changes in regulations, where the license change 
results in very minor changes to facility operations clearly in 
keeping with the regulations. The requested amendments are similar 
to example vii in that they result in minor changes to plant 
surveillance requirements and are consistent with the existing NRC 
position and guidance contained in NUREG-1366 and Generic Letter 93-
05, as well as NUREG-1431. While the issuance of NUREG-1366 and 
Generic Letter 93-05, as well as NUREG-1341 does not constitute a 
change in existing regulations, it nevertheless establishes the NRC 
staff's position concerning the acceptability of decreasing the 
surveillance frequency of AFW pumps from monthly to quarterly and 
concerning the acceptability of adopting all or part of the new STS. 
The requested amendments are consistent with the position of NUREG-
1366 and with the guidance of Generic Letter 93-05, as well as with 
NUREG-1431.

Criterion 1

    The requested amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Decreasing the frequency of AFW pump testing as specified 
in TS from monthly to quarterly will have no impact upon the 
probability of any accident, since the AFW pumps are not accident 
initiating equipment.
    Also, since... McGuire's AFW pump performance histories support 
making the proposed change, system response following an accident 
will not be adversely affected. Therefore, the requested amendments 
will not result in increased accident consequences. ***Incorporating 
the new STS note will only serve to clarify when the turbine-driven 
pump is required to be tested and will not have any impact upon 
either the probability or consequences of any accident. The pump 
will still be tested as before and its acceptance criteria will be 
unaffected.

Criterion 2

    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated above, the AFW pumps are not accident 
initiating equipment. No new failure modes can be created from an 
accident standpoint. The plant will not be operated in different 
manner. ***Incorporating the clarifying note from the new STS will 
not result in any new accident sequences, since plant operation will 
be unaffected.

Criterion 3

    The requested amendments will not involve a significant 
reduction in a margin of safety. Plant safety margins will be 
unaffected by the proposed changes. The AFW pumps will still be 
capable of fulfilling their required safety function, since plant 
operating experience supports the proposed change. The availability 
of the AFW pumps will be increased as a result of the proposed 
amendments because they will not have to be made unavailable for 
testing as frequently. Finally, the proposed amendments are 
consistent with the NRC position and guidance set forth in NUREG-
1366 and Generic Letter 93-05. ***Incorporating the note from the 
new STS will not impact any safety margins.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The proposed technical specification amendment has been reviewed 
against the criteria of 10 CFR 51.22 for environmental 
considerations. The proposed amendment does not involve a 
significant hazards consideration, nor increase the types and 
amounts of effluents that may be released offsite, nor increase 
individual or cumulative occupational radiation exposures. 
Therefore, the proposed amendment meets the criteria given in 10 CFR 
51.22(c)(9) for a categorical exclusion from the requirement for an 
Environmental Impact Statement.

    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: Atkins Library, University of 
North Carolina, Charolotte (UNCC Station), North Carolina 28223.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: June 13, 1994.
    Description of amendment request: The proposed amendment would 
modify Clinton Power Station Technical Specification 3/4.6.1.8, 
``Containment Building Ventilation and Purge Systems,'' which includes 
a requirement to perform a leak rate measurement at least once per 92 
days on each 36-inch supply and exhaust containment ventilation 
isolation valve with a resilient seal. The proposed modification would 
require a leak test at least once per 18 months provided the valves 
remain closed during that period.
    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 change in plant 
design. Failure of or leakage through a containment barrier cannot 
itself create an accident; therefore, this change would not increase 
the probability of any accident previously evaluated. Failure of or 
leakage through a containment barrier can, however, increase the 
consequences of those accidents previously evaluated. The proposed 
change merely revises the frequency at which the local leak rate 
test is performed on the containment building HVAC 36-inch supply 
and exhaust penetrations. The containment isolation valves for these 
penetrations are normally only opened during refueling outages. The 
stroke testing for the isolation valves has been changed to a cold 
shutdown frequency and, as a result, there is no mechanism present 
to degrade the seals and cause increased leakage through the 
penetration. Based on past penetration leak rate measurements, it 
has been determined that leak rate testing on an 18-month frequency 
is sufficient to identify seal degradation if the valves are not 
opened. However, should the valves be opened during the 18-month 
interval, the proposed change would require that a leak rate test be 
performed within 92 days. This will ensure that the leak rate for 
the given penetration has not exceeded the specified limit as a 
result of stroking the valve. Penetration leakage will continue to 
be measured at sufficient intervals to identify seal degradation in 
the 36-inch containment isolation valves. In addition, the same 
leakage limits will be imposed. Therefore, the proposed change will 
not result in a significant increase in the probability or the 
consequences of any accident previously evaluated.
    (2) This request does not result in any change to the plant 
design nor does it involve a change in current plant operation. The 
proposed change will not change the design basis for the valves 
being leak tested. The valves will continue to be verified to meet 
the required leak rate and the safety function of the subject valves 
remains unchanged. Furthermore, any potential leakage through the 
containment building HVAC 36-inch supply exhaust and supply 
penetrations cannot create an accident. As a result, the proposed 
change cannot create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) The only margin of safety that could potentially be impacted 
by the proposed change to the surveillance requirement frequency is 
the margin concerning the offsite dose consequences of postulated 
accidents (which is directly related to the containment leak rate). 
As discussed above, this request does not result in a significant 
increase in the consequences of any accident previously evaluated. 
It has been demonstrated that the penetration leakage does not 
change appreciably when the valves are not stroked. Therefore, since 
the valves are normally only opened during refueling outages, 
leakage through the penetrations is not expected to change during 
the proposed 18-month interval between leak rate tests. Should the 
valves be opened during the 18-month interval, a local leak rate 
test will be performed within 92 days. The proposed leak rate test 
frequency will provide sufficient indication of seal degradation to 
allow the opportunity for repair before gross leakage failures 
develop. In addition, the proposed change involves no change to the 
currently established leak rate test acceptance criteria. As a 
result, the proposed changes do not result in 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur, 
IL 62525.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: August 12, 1994.
    Description of amendment request: The proposed amendment would 
modify Clinton Power Station Technical Specification 3/4.6.2.2, 
``Drywell Bypass Leakage,'' to allow drywell by pass leakage rate tests 
(DBLRTs) to be performed at intervals as long as ten years based on the 
demonstrated performance of the drywell structure. DBLRTs are currently 
required to be performed once every 18 months.
    Basis of 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 change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated.
    The proposed change potentially affects the leak tight integrity 
of the drywell, a structure used to mitigate the consequences of a 
loss of coolant accident (LOCA). The function of the drywell is to 
channel the steam released from the LOCA through the suppression 
pool, limiting the amount of steam released to the primary 
containment atmosphere. This limits the containment pressurization 
due to the LOCA. The leakage of the drywell is limited to ensure 
that the primary containment does not exceed its design limits of 
185  deg.F and 15 psig. Because the proposed change does not alter 
the plant design, only the frequency of measuring the drywell 
leakage, the proposed change does not directly result in an increase 
in the drywell leakage. However, decreasing the test frequency can 
increase the probability that a large increase in drywell bypass 
leakage could go undetected for an extended period of time. There 
are several potential sources of steam bypass leakage paths. These 
include potential cracks in drywell concrete structure, the drywell 
vacuum breakers, and various penetrations through the drywell 
structure. Based on the results of the structural integrity test 
conducted as part of the preoperational test program, additional 
cracking of the drywell is not expected during the remaining life of 
the plant. Ventilation and piping penetrations (including the 
drywell vacuum breaker penetrations) are designed to ASME Code Class 
2 and Seismic Category 1 requirements. These penetrations are 
designed with two isolation valves in series with one valve in the 
drywell and another either outside primary containment or in the 
wetwell. High energy lines that extend into the wetwell, such as the 
main steam lines and feedwater lines, are encapsulated by guard 
pipes to direct energy to the drywell in case of a piping rupture. 
Electrical penetrations are sealed with a high strength/density 
material that will prevent leakage as well as provide radiation 
shielding. Operational experience has shown that the leak tightness 
of the drywell has maintained well below the allowable leakage 
limits. In fact, the calculated drywell bypass leakage area is of 
such a small magnitude that containment design pressure could not be 
exceeded even if containment spray and heat sinks were not 
available. The technical specification limit of 10% of the maximum 
allowable leakage path area provides margin for degradation. Drywell 
performance data to date suggest that drywell degradation, even 
during a ten-year interval between tests, will not exceed this 
margin.
    Further, an analysis was conducted to determine the potential 
risk to the public from unacceptable drywell bypass leakage going 
undetected as a result of the proposed change. Based on this 
analysis, under several different accident scenarios, the risk of 
radioactivity release from containment was found to be negligible, 
about 10-9 per year.
    Based on the above, Illinois Power has concluded that the 
proposed change will not result in a significant increase in the 
consequences of any accident previously evaluated.
    (2) The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. Thus, the proposed change cannot create 
the possibility of an accident not previously evaluated.
    (3) The proposed change only affects the frequency of measuring 
the drywell leakage and does not change the bypass leakage limit for 
the drywell. However, the proposed change can increase the 
probability that a large increase in drywell bypass leakage could go 
undetected for an extended period of time. Operational experience 
has shown that the leak tightness of the drywell has been maintained 
well below the allowable leakage limits. In fact, the calculated 
drywell bypass leakage area is of such a small magnitude that 
containment design pressure could not be exceeded even if 
containment spray and heat sinks were not available. Further, an 
analysis was conducted to determine the potential risk to the public 
from the proposed change. Based on this analysis, under several 
different accident scenarios, the risk of radioactivity release from 
containment was found to be negligible, about 10-9 per year. As 
a result, Illinois Power has concluded that the proposed change will 
not result in 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur, 
IL 62525.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: August 12, 1994.
    Description of amendment request: The proposed amendment would 
modify Clinton Power Station Technical Specifications 3/4.3.1, 
``Reactor Protection System Instrumentation,'' 3/4.3.2, ``Containment 
and Reactor Vessel Isolation Control System,'' 3/4.3.3, ``Emergency 
Core Cooling System Actuation Instrumentation,'' 3/4.3.4.2, ``End-of-
Cycle Recirculation Pump Trip System Instrumentation,'' 3/4.3.5, 
``Reactor Core Isolation Cooling System Actuation Instrumentation,'' 3/
4.4.2.1, ``Safety/Relief Valves,'' and 3/4.4.2.2, ``Safety/Relief 
Valves Low-Low Set Function.'' These technical specifications contain 
requirements to perform manual testing of the associated solid-state 
logic at least once every four fuel cycles. This testing is in addition 
to the automatic testing performed by the self-test system. Due to the 
negative impact on plant safety caused by the need to remove systems 
from service to prevent unwanted actuations and the increased potential 
for unintended equipment actuation during manual testing, Illinois 
Power is proposing that the requirement to perform manual testing of 
the solid-state logic independently from the self-test system be 
eliminated.
    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 change to the plant 
design. The proposed change involves only testing of the solid-state 
Nuclear Systems Protection System (NSPS) utilizing the self-test 
system (STS). As identified in Supplement No. 2 to the Clinton Power 
Station (CPS) Safety Evaluation Report (SSER 2), use of the STS to 
perform certain surveillance testing required by the plant Technical 
Specifications is acceptable. However, as noted in SSER 6, portions 
of the NSPS logic must also be manually tested during each refueling 
outage independently from the STS such that all NSPS trip/actuation 
functions are tested independently of the STS at least once every 
four fuel cycles. The change proposed in this request consists of 
the elimination of this manual testing of the NSPS logic 
independently from the STS.
    As identified in SSER 6, the purpose of the currently required 
manual tests is to provide a means to verify operability of the NSPS 
functional circuits independent of the STS, and thereby (1) detect 
any failures undetected by the STS and take corrective action to 
restore proper operation of the NSPS and STS and (2) assuming that 
no additional failures beyond those identified by the STS are 
detected during independent logic testing, confirm the validity of 
the STS test results. The ability of the STS to detect functional 
failures of the NSPS logic as designed was verified as part of 
General Electric's independent design verification. In addition, 
this capability was also verified by testing as part of the 
preoperational test program as described in Section 14.2.12.1.62 of 
the CPS Updated Safety Analysis Report (USAR) and during testing 
performed during the first four refueling outages in accordance with 
the current Technical Specification testing requirements. Thus, 
Illinois Power (IP) has concluded that all functional failures 
undetectable by the STS have been identified and the aforementioned 
objectives have been satisfied.
    Since the proposed change does not alter the plant design or 
operation, it cannot increase the probability of any accident 
previously evaluated. This proposed change does involve the NSPS 
logic which is utilized to actuate systems needed to mitigate the 
consequences of accidents previously evaluated, however, the 
proposed change merely eliminates the currently required manual 
testing independently from the STS once every four fuel cycles. 
Since the proposed change does not alter the NSPS logic, does not 
impact operation of the STS, and continues to require adequate 
testing of the NSPS logic on a frequency sufficient to maintain the 
operability of the associated NSPS logic, the proposed change cannot 
impact the reliability of the associated actuation instrumentation 
and therefore cannot increase the consequences of any accident 
previously evaluated.
    (2) Adequate testing of the NSPS logic will continue to be 
required. The proposed change continues to allow us of the STS in 
performing surveillance tests as documented by the NRC in SSER 2. 
However, additional manual tests independently from the STS will no 
longer be required. Since the proposed change does not add 
additional testing configurations or operating modes nor does it 
alter the plant design, it will not introduce any new failure modes. 
Thus, this proposed change cannot create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Performance of the currently required manual testing of the 
NSPS logic independently from the STS system involves actuation 
logic for the reactor protection system, emergency core cooling 
systems (ECCS), reactor core isolation cooling system, automatic 
depressurization system, nuclear steam supply shutoff system, and 
the residual heat removal (RHR) system, including the shutdown 
cooling mode of operation. Performance of these tests requires these 
systems to be disabled to prevent unwanted system actuations.Thus, 
these systems are rendered inoperable during this testing. This also 
results in extensive temporary reconfiguration of systems and 
actuation instrumentation, including logic card removal, 
installation of signal simulators, disconnecting load drivers, etc. 
The removal of these safety systems from service to perform this 
testing results in reduced availability of RPS, ECCS, and RHR 
shutdown cooling systems during the plant outage. In addition to the 
intentional disabling of equipment to perform this testing, this 
testing has in the past led to safety system unavailability due to 
equipment damage caused by bending pin connectors and burning 
circuit cards out due to electrical shorting. Further, CPS has 
experienced unintentional equipment actuations resulting in 
unnecessary challenges to safety systems and the need to file 
licensee event reports with the NRC. IP has concluded that 
elimination of this currently required manual testing of the NSPS 
will have a positive impact on overall plant safety.
    The only margin of safety that could be negatively impacted by 
this proposed change is the potential for a functional failure in 
the NSPS logic going undetected. The manual tests proposed for 
deletion are only required to be performed at least once every four 
fuel cycles. In addition, the ability of the STS to detect 
functional failures of the NSPS logic as designed was verified as 
part of General Electric's independent design verification. Further, 
this capability was also verified as part of the preoperational test 
program as described in Section 14.2.12.1.62 of the CPS Updated 
Safety Analysis Report (USAR) and subsequently during testing 
performing during the first four refueling outages. These manual 
tests did not identify any functional failures of the NSPS logic 
which would be expected to be detected by the STS per its design. 
All other functional failures undetectable by the STS have been 
identified as Untested Islands (UTIs). Thus, the original objectives 
of this testing have been satisfied.
    The proposed change will continue to require testing at 
refueling outage intervals. As identified in USAR Section 
7.2.1.1.4.8 and SSER 6, circuits which are not capable of being 
tested by the STS are identified as UTIs. General Electric and IP 
have identified, via analysis and manual testing, all UTIs in the 
functional NSPS logic and have established procedures for testing 
these UTIs. Periodic verification of the operability of these UTIs 
will continue to be performed at the frequencies recommended by the 
manufacturer as accepted by the NRC in SSER 6. These frequencies 
were established by the manufacturer based on mean time between 
failure analyses for the components in the associated circuits and 
may be as long as six years. Based on the above, the functional 
operability of the NSPS logic is adequately assured.
    From the above, IP has concluded that the proposed change will 
result in a net increase in the overall 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th St., Decatur, 
IL 62525.
    NRC Project Director: John N. Hannon.

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

    Date of amendment requests: April 6, 1994.
    Description of amendment requests: The proposed amendments would 
remove the license condition reference to Table 1 of the Fire 
Protection Safety Evaluation Report (SER) for Cook Nuclear Plant, 
approved on June 4, 1979, and issued under a cover letter dated July 
31, 1979. Table 1 of the 1979 First Protection SER is a schedule for 
completion of 23 modifications which have since been completed. Three 
of the modifications (Nos. 7C, 9, & 20) have been changed since the 
1979 Fire Protection SER.
    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

    This update of the fire protection modifications contained in 
Table 1 of the July 31, 1979, fire protection SER does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed as follows:
    Item 7C--The use of an air compressor rather than the cascade 
recharging stations does not adversely impact the ability to supply 
breathing air for the fire brigade and is acceptable under BTP APCSB 
9.5-1.
    Item 9--the use of unrated metal hatches has been accepted by 
the NRC in a SER date June 17, 1988.
    Item 20--The new fire pumps have already been accepted by the 
NRC in the SER dated March 31, 1993, Amendment Nos. 171 and 154. The 
retirement of the old screenhouse diesel driven fire pumps is based 
on the availability of another supply of ``backup'' fire suppression 
water.

Criterion 2

    This update of the fire protection modifications contained in 
Table 1 of the July 31, 1979, fire protection SER does not create 
the possibility of a new or different kind of accident from any 
accident previously evaulated as follows:
    Item 7C--The use of the an air compressor rather than the 
cascade recharging stations does not adversely impact the ability to 
supply breathing air for the fire brigade and is acceptable under 
BTP APCSB 9.5-1.
    Item 9--The use of unrated metal hatches has been accepted by 
the NRC in a SER dated June 17, 1988.
    Item 20--The new fire pumps have already been accepted by the 
NRC in the SER dated March 31, 1993, Amendment Nos. 171 and 154. The 
retirement of the old screenhouse diesel driven fire pumps is based 
on the availability of another supply of ``backup'' fire suppression 
water.

Criterion 3

    This update of the fire protection modifications contained in 
Table 1 of the July 31, 1979, fire protection SER does not involve a 
significant reduction in a margin of safety as follows:
    Item 7C--The use of an air compressor rather than the cascade 
recharging stations does not adversely impact the ability to supply 
breathing air for the fire brigade and is acceptable under BTP APCSB 
9.5-1. Filled spare breathing air bottles are on-site and the local 
municipal department will provide assistance as needed.
    Item 9--The use of unrated metal hatches has been accepted by 
the NRC in a SER dated June 17, 1988.
    Item 20--The new fire pumps have already been accepted by the 
NRC in the SER dated March 31, 1993, Amendment Nos. 171 and 154. The 
retirement of the old screenhouse diesel driven fire pumps is based 
on the availability of another supply of ``backup'' fire suppression 
water.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: July 19, 1994
    Description of amendment request: The proposed amendments would 
revise the technical specifications by removing the specific scheduling 
requirements for Types A, B, and C tests and replacing these 
requirements with a requirement to perform Types A, B, and C testing in 
accordance with Appendix J to 10 CFR 50.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1

    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed changes to the T/Ss do not affect the 
assumptions, parameters, or results of any UFSAR accident analysis. 
The proposed changes do not modify the response of the containment 
during a design basis accident. The proposed amendment does not add 
or modify any existing equipment. The proposed Types A, B, and C 
testing schedules will be consistent with Appendix J to 10 CFR 50. 
Based on these considerations, it is concluded that the changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2

    The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The proposed 
changes only remove the restrictive schedular requirements for 
conducting Type A testing from the T/Ss and substitute the schedule 
specified in Appendix J to 10 CFR 50. For Types B and C testing, the 
schedular requirements are removed from T/Ss because they are 
already specified in Appendix J to 10 CFR 50. Thus, it is concluded 
that the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3

    The margin for safety presently provided is not reduced by the 
proposed change in the schedular requirements for Type A tests. 
Types B and C schedular requirements are not changed by removing 
them from T/Ss. Although the changes allow more flexibility in 
scheduling Type A tests, the proposed amendment continues to ensure 
reactor containment system reliability by periodic testing in full 
compliance with 10 CFR 50, Appendix J. Based on these 
considerations, it is concluded that the changes do not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: July 26, 1994.
    Description of amendment request: The proposed amendments would 
modify the technical specifications such that the requirement to 
measure the moderator temperature coefficient near the end of the cycle 
will become conditional. The test will not be performed if specified 
core performance benchmark criteria are met for the operating cycle, 
and the revised predicted moderator temperature coefficient is less 
negative than the moderator temperature coefficient surveillance limit 
presented in the Core Operation Limits Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1

    Does the EOL MTC measurement conditional exemption involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    No. The conditional exemption of the most negative MTC 
measurement does not change the most negative MTC surveillance and 
LCO limits in the T/Ss. Since these MTC values are unchanged, and 
since the basis for the derivation of these values from the safety 
analysis MDC is unchanged, the constant MDC assumed for the FSAR 
safety analyses will also remain unchanged. Therefore, no change in 
the modeling (i.e., probabilities) of the accident analysis 
conditions or response is necessary in order to implement the change 
to the conditional exemption methodology. In addition, since the 
constant MDC assumed in the safety analyses is not changed by the 
conditional exemption of the most negative MTC surveillance 
measurement, the consequences of an accident previously evaluated in 
the FSAR are not increased. The dose predictions presented in the 
FSAR for a SGTR remain valid such that more severe consequences will 
not occur. Additionally, since mass and energy releases for LOCA and 
steamline break are not increased as a result of the unchanged MDC, 
the dose predications for these events presented in the FSAR also 
remain bounding.

Criterion 2

    Does the EOL MTC measurement conditional exemption create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    No. Since the EOL MTC is not changed by the conditional 
exemption methodology of Reference 1, the possibility of an accident 
which is different than any already evaluated in the FSAR has not 
been created. No new or different failure modes have been defined 
for any system or component nor has any new limiting single failure 
been identified. Conservative assumptions for MDC have already been 
modeled in the FSAR analyses. These assumptions will remain valid 
since the conditional exemption methodology documented in Reference 
1 does not change the safety analysis MDC nor the T/S values of the 
MTC.

Criterion 3

    Does the EOL MTC measurement conditional exemption involve a 
significant reduction in a margin of safety?
    No. The evaluation of the conditional exemption methodology 
documented in Reference 1 has taken into account the applicable Cook 
Nuclear Plant Units 1 and 2 T/Ss and has bounded the conditions 
under which the specifications permit operation. The applicable T/Ss 
are surveillance 4.1.1.4, and reference ``e'' is added to the list 
of references in Specification 6.9.1.11.2. An additional 
specification 6.9.1.12 is added to define the requirements for the 
``Most Negative Moderator Temperature Coefficient Limit Report,'' 
which is described in Appendices A, C, and D of Reference 1. The 
COLR has also been modified as described in Appendix B of Reference 
1. The analyses which support these T/Ss have been evaluated. The 
results, as presented in the FSAR, remain bounding since the MDC 
assumed in the safety analyses and the LCO and surveillance 
requirement MTCs in the T/Ss remain unchanged. Therefore, the margin 
of safety, as defined in the bases to these T/Ss, is not reduced.

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

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

    Date of amendment request: August 26, 1994
    Description of amendment request: The proposed amendment to 
Technical Specification 4.3.3.c(1) would allow a one-time extension of 
the 10 year service period for the Primary Containment Integrated 
Leakage Rate (Type A) Test. Specifically, the proposed one-time only 
change would extend the 10-year service interval requirement for 
performance of the Type A test to correspond with the end of the 
current inservice inspection interval (ISI). The interval extension 
would avoid the necessity of performing an additional Type A test only 
22 months after the previous test. This would result in an extension of 
the second interval from 10 years to approximately 14 years and 46 
months between the second and third test.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed extension of the Type A test 10-year service 
interval does not increase the chances for a previously analyzed 
accident to occur. Containment integrity is required for the 
mitigation of accident consequences. Furthermore, containment 
leakage is not the precursor to any analyzed event. Extension of the 
Type A test surveillance interval will not affect the containment's 
ability to maintain leakage below that assumed in the safety 
analysis. The previous Type A test was completed successfully and no 
plant modifications have been made or are planned (other than those 
that require Type B or C testing) since the last test which could 
directly affect the test results. Type B and C testing of individual 
penetrations has been satisfactory and will continue to be performed 
in accordance with the Technical Specifications. There have been no 
pressure or temperature excursions in the containment which could 
have adversely affected containment integrity. Hence, the ability of 
containment to maintain leakage within the Type A test limits be 
maintained.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed one-time extension of the Type A test 10-year 
service interval will not affect the test methodology or acceptance 
criteria nor does it alter the physical containment structure or 
boundary in any way. There will be no addition or removal of plant 
hardware. No new plant operating modes are being introduced. Results 
of the previous Type A tests are well below allowable limits, and 
there have been no plant modifications (other than those that 
require Type B or C testing) since the last test nor are any 
planned, that could directly impact the previous Type A test 
results.
    Therefore, the proposed change will not create the possibility 
of a new or different accident from any previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    Safety margins are established through the Nine Mile Point Unit 
1 safety analyses as reflected in the Technical Specification 
Limiting Conditions for Operation. Containment leak rates assumed in 
the safety analyses are not increased by the proposed change to the 
Type A test 10-year service interval. The acceptance criteria which 
must be met to verify that leak rates remain within assumed values 
will not be changed.
    Although the test frequency will be relaxed for the one-time 
extension, no plant modifications have been made or are planned 
which would invalidate the last Type A leak test results which 
confirm acceptable containment integrity. Furthermore, Type B and C 
testing of individual penetrations has been satisfactory and will 
continue to be performed in accordance with the Technical 
Specifications to assure that containment integrity is maintained.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of a safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, NY 
13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael J. Case.

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

    Date of amendment request: August 26, 1994.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) Section 3/4.6.1.3, ``Primary Containment 
Air Locks.'' Specifically, TS 3/4.6.1.3 would be revised to allow 
continued plant operation if an interlock becomes inoperable as long as 
an operable door is locked shut and periodically checked as being 
locked shut.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The primary containment and containment air locks are not 
initiators or precursors to an accident. Primary containment 
integrity ensures that the release of radioactive materials from the 
containment will be restricted to those leakage paths and associated 
leak rates assumed in the accident analyses. Therefore, the proposed 
changes to the air lock ACTION statements cannot affect the 
probability of a previously evaluated accident.
    The purpose of a primary containment air lock interlock is to 
allow only one door to be opened at a time in each penetration. This 
provision ensures that a gross breach of primary containment does 
not exist when primary containment is required to be operable. 
Closure of a single door in each air lock is sufficient to provide a 
leak tight barrier following postulated events. If an air lock 
interlock is inoperable, the proposed ACTION requires that an 
operable door be locked shut and periodically verified locked and 
shut. This assures that at least one air lock door is closed, which 
provides the function of the interlock, thereby assuring containment 
integrity is maintained. Therefore, the proposed change will not 
significantly increase the consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes to the primary containment air lock 
specification will allow contained plant operation with an 
inoperable interlock as long as an operable door is locked closed 
and periodically verified to be locked closed. The changes do not 
introduce any new accident precursors and do not involve any 
alterations to plant configurations which could initiate a new or 
different kind of accident. The change provides an alternate means 
of ensuring that only one primary containment air lock door is 
opened at a time in each penetration. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed changes to the primary containment air lock ACTION 
statements will not affect the ability of the containment to respond 
to an accident and limit releases to within 10 CFR Part 100 and GDC 
[General Design Criterion] 19 guidelines. The changes do not affect 
the design or performance characteristics of the containment or 
containment air locks but simply provide an alternative means of 
ensuring that only one air lock door is opened at a time. 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 request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael J. Case.

 Niagara Mohawk Power Corporation, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New York

    Date of amendments request: June 9, 1994.
    Description of amendments request: The proposed change would revise 
Section 3.4, ``Response Force Capabilities,'' of the Nine Mile Point 
Nuclear Station Physical Security Plan with regard to the number of 
armed Security Force Members comprising the Response Force for each 
shift.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation of Nine Mile Point Units 1 and 2 in accordance 
with the proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment to the Physical Security Plan involves a 
change in the armed response force size for Units 1 and 2. Nuclear 
Security's ability to meet the response requirements of 10 CFR 
[subsection] 73.55(h) will not be significantly affected as was 
determined utilizing NUREG-0907 while taking into consideration 
several Security Program enhancements implemented since 1986 when 
response force size was last reviewed.
    The change does not affect the design, function, operation, 
maintenance or testing of structures, systems and components at Unit 
1 or Unit 2. It does not affect safety analysis or any Technical 
Specification that preserves a safety analysis assumption. 
Therefore, the proposed amendment will neither increase the 
probability nor the consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Units 1 and 2 in accordance 
with the proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment to the Physical Security Plan involves a 
change in the size of the response force for Units 1 and 2. Design, 
function, operation, maintenance and testing of structures, systems 
and components at Units 1 and 2 are not affected. Also, as 
demonstrated using NUREG-0907, Nuclear Security will continue to 
meet the requirements of 10 CFR [subsection] 73.55(h). Therefore, 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The operation of Nine Mile Point Units 1 and 2 in accordance 
with the proposed amendment will not involve a significant reduction 
in a margin of safety.
    The proposed amendment to the Physical Security Plan involves a 
change in the size of the armed response force for Units 1 and 2. 
The proposed change has no impact on the physical design of the 
plants nor on the function or operation of their structures, systems 
and components. The proposed amendment does not impact and therefore 
does not reduce the margin of safety as defined in the basis for any 
Unit 1 or Unit 2 Technical Specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael J. Case.

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

    Date of amendment request: September 1, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications concerning the Reactor Coolant System 
Volume (RCS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    * * * The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Section 5.4.2 of the Millstone Unit No. 2 
Technical Specifications revises the total RCS water and steam 
volume to reflect the installation of new steam generators. Section 
5.0 of the Millstone Unit No. 2 Technical Specifications delineates 
design features for Millstone Unit No. 2. The proposed change does 
not modify a limiting condition for operation, action statement, or 
surveillance requirement. Additionally, the proposed change does not 
revise the manner in which the plant is operated. It simply revises 
the value delineated for the total RCS water and steam volume.
    The steam generator replacement modifications were addressed in 
a plant design change request performed in accordance with the 
requirements of 10 CFR 50.59. NNECO concluded that the replacement 
of the steam generators did not involve an unreviewed safety 
question. This proposed change is a result of those modifications.
    Based on the above, the proposed change to Section 5.4.2 of the 
Millstone Unit No. 2 Technical Specifications does 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 analyzed.
    The proposed change to Section 5.4.2 of the Millstone Unit No. 2 
Technical Specifications simply revises the total RCS water and 
steam volume to reflect the installation of new steam generators. It 
does not modify the manner in which any plant equipment or systems 
are operated. Thus, the proposed change to Section 5.4.2 of the 
Millstone Unit No. 2 Technical Specifications 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.
    The proposed change to Section 5.4.2 of the Millstone Unit No. 2 
Technical Specifications revises the total RCS water and steam 
volume to reflect the installation of new steam generators. This 
proposal simply reflects the slight change in volume due to the 
replacement of the steam generators. The steam generator replacement 
modifications were addressed in a plant design change request 
performed in accordance with the requirements of 10 CFR 50.59. NNECO 
concluded that the replacement of the steam generators did not 
involve an unreviewed safety question.
    Based on the above, the proposed change to Section 5.4.2 of the 
Millstone Unit No. 2 Technical Specifications 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 Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: August 22, 1994.
    Description of amendment request: The proposed amendment would 
change Technical Specifications 3/4.1.3: to extend the scram discharge 
volume (SDV) vent or drain valve restoration time from the current time 
period of 24 hours to 7 days; to permit the SDV vent and drain valves 
operability check to be performed at shutdown conditions instead of at 
least once per 18 months; and to delete the SDV float switch response 
surveillance.
    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:

    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to Technical Specification 3.1.3.1.d, which 
extends the restoration time for one inoperable vent and/or drain 
valve from 24 hours to 7 days, does not significantly increase the 
probability or consequences of an accident previously evaluated. A 7 
day restoration time, as identified in NUREG 1433, is reasonable 
given the level of redundancy in the lines and the low probability 
of a scram occurring while the valve is inoperable and the line is 
not isolated. If the inoperable valve fails open, the redundant 
valve in the line allows for leakage from the CRD [control rod 
driveline] to be drained out and also allows for the line to be 
isolated if necessary. If the valve fails closed, the line becomes 
isolated. However, float switches and pressure sensors will notify 
operators of water buildup in the instrument volume. A review of the 
surveillance data indicates the vent and drain valves rarely fail 
the initial operability test and require rework. The low failure 
rate combined with the redundancy of the valves makes for a highly 
reliable system. Therefore, the proposed change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Changing Surveillance Requirement 4.1.3.1.4.a from requiring 
demonstration of the SDV vent and drain valve operability during a 
scram at less than or equal to 50% rod density to a requirement to 
perform the testing at shutdown conditions does not significantly 
increase the probability or consequences of an accident previously 
evaluated. The purpose of the 50% rod density requirement is to 
provide a test environment having typical reactor coolant pressure 
and temperature conditions. However, the closure time of the vent 
and/or drain valves is not affected by pressure and any variations 
due to temperature are relatively insignificant. Therefore, testing 
from shutdown conditions ensures the safety functions of the vent 
and drain valves are met. Also, the proposed change does not affect 
system design or operation. Therefore, the proposed change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The deletion of Surveillance Requirement 4.1.3.1.4.b, requiring 
proper float switch response by verifying float switch actuation 
after a scram, does not significantly increase the probability or 
consequences of an accident previously evaluated. Design changes 
eliminated the high differential pressure experienced by the float 
switches after a scram and provide redundant level measuring 
instrumentation. Differential pressure gauges were added in addition 
to the float switches to provide the RPS [reactor system protection] 
logic with a diverse and redundant means of measuring SDIV [SDV] 
level. The changes have resulted in no crushed ball floats at SSES 
after a scram. The proposed change will have a negligible effect 
[on] the reliability of the system and therefore, does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change to Technical Specification 3.1.3.1.d, which 
extends the restoration time for one inoperable vent and/or one 
inoperable drain valve from 24 hours to 7 days, does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Extending the restoration time to 7 days does 
not change the design purpose or operation of the SDV valves. 
Therefore, the change is bounded by the existing accident analysis.
    Changing Surveillance Requirement 4.1.3.1.4.a from requiring 
demonstration of the SDV vent and drain valve operability during a 
scram at less than or equal to 50% rod density to a requirement to 
perform the testing at shutdown conditions does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. Performing the surveillance from shutdown 
conditions, as recommended in NUREG 1366, ensures that the 
operability of the SDV is maintained. No new failure modes are 
introduced by the change and the change is bounded by the existing 
accident analysis.
    The deletion of Surveillance Requirement 4.1.3.1.4.b, requiring 
proper float switch response by verifying float switch actuation 
after a scram, does not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The role of 
the ball floats to sense increases in SDV water level is not 
affected by the change. In the unlikely event that crushing of the 
ball float were to occur, redundant level measuring devices would 
maintain system function and the existing 92 day surveillance 
activity would identify a ball float failure.
    III. This change does not involve a significant reduction in a 
margin of safety.
    The proposed change to Technical Specification 3.1.3.1.d, which 
extends the restoration time for one inoperable vent and/or one 
inoperable drain valve from 24 hours to 7 days, does not involve a 
significant reduction in a margin of safety. A 7 day completion 
time, as identified in NUREG 1433, is reasonable given the level of 
redundancy in the lines and the low probability of a scram occurring 
while the valve is inoperable and the line is not isolated. Also, a 
separate Technical Specification (3.1.3.1.e) addresses the potential 
for two inoperable vent or drain valves in the same line.
    Changing Surveillance Requirement 4.1.3.1.4.a from requiring 
demonstration of the SDV vent and drain valve operability during a 
scram at less than or equal to 50% rod density to a requirement to 
perform the testing at shutdown conditions does not involve a 
significant reduction in a margin of safety. The change maintains 
the intent of S.R. 4.1.3.1.4.a by performing equivalent testing at 
the same frequency. The change increases the margin of safety by 
eliminating potential future scrams taken to meet S.R. 4.1.3.1.4.a., 
thus reducing the potential for safety challenges.
    The deletion of Surveillance Requirement 4.1.3.1.4.b, requiring 
proper float switch response by verifying float switch actuation 
after a scram does not involve a significant reduction in a margin 
of safety. Design changes have enhanced the SDV design and provide a 
redundant and diverse means of monitoring SDV level. Also, a 92-day 
surveillance (S.R. 4.3.1.1) ensures the operability of the float 
switch.

    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Mohan Thadani (Acting).

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

    Date of amendment requests: December 30, 1993, June 3, 1994, and 
August 25, 1994.
    Description of amendment requests: The licensee proposes to replace 
the current Technical Specifications (TS) with a set of TS based on the 
CE Owners Group Improved Standard Technical Specifications issued by 
the NRC staff as NUREG-1432 in September 1992. The adoption of Owners 
Group-approved TS is part of an industry-wide initiative to standardize 
and improve TS. San Onofre, Units 2 and 3, is the lead plant for 
adoption of the CE Owners Group standardized TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates requirements and surveillances for 
structures, systems, components or variables which did not meet the 
criteria for inclusion in Technical Specifications. The affected 
structures, systems, components or variables are not assumed to be 
initiators of analyzed events and are not assumed to mitigate 
accident or transient events. The requirements and surveillances for 
these affected structures, systems, components or variables will be 
relocated from the Technical Specifications to the Licensee 
Controlled Specifications or the UFSAR [updated final safety 
analysis report]. These operability requirements and surveillances 
will continue to be maintained pursuant to 10 CFR 50.59. In 
addition, the affected structures, systems, components or variables 
are addressed in existing surveillance procedures which are 
controlled by 10 CFR 50.59. The reformatting and rewording process 
involves no technical changes to plant design or operations. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will not impose any different 
requirements and adequate control of information will be maintained. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analyses assumptions. In addition, 
the affected structure, system, component or variable requirements 
and surveillances are the same as the existing Technical 
Specifications. Since any future changes to these requirements in 
the Licensee Controlled Specifications or UFSAR in the surveillance 
procedures will be evaluated per the requirements of 10 CFR 50.59, 
no significant reduction in a margin of safety will be allowed. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: Theodore R. Quay.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: August 17, 1994.
    Description of amendments request: The proposed amendment would 
modify Surveillance Requirement 4.3.1.2, Reactor Trip System 
Instrumentation, and Surveillance Requirement 4.3.2.1, Engineered 
Safety Feature Actuation System Instrumentation, to eliminate the 
periodic pressure sensor response time testing requirements. The 
Surveillance Requirements would indicate that the total channel 
response time will be periodically ``verified'' instead of ``tested''.
    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 license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same RTS 
and ESFAS instrumentation is being used; the time response 
allocations/modeling assumptions in the FSAR Chapter 15 analyses are 
still the same; only the method of verifying time response is 
changed. Periodic calibration of these pressure and differential 
pressure instruments will detect significant degradation in the 
sensor response characteristic and assure equipment operability. 
Flushing of selected sensing lines during each refueling outage as 
recommended by NUREG/CR5851, ``Long Term Performance and Aging 
Characteristics of Nuclear Plant Pressure Transmitters,'' will 
mitigate sensing line response time degradation due to blockage that 
noise analysis testing techniques would have previously detected. 
The proposed change will not modify any system interface and could 
not increase the likelihood of an accident since these events are 
independent of this change. The proposed activity will not change, 
degrade or prevent actions or alter any assumptions previously made 
in evaluating the radiological consequences of an accident described 
in the FSAR. Therefore, the proposed amendment does not result in 
any increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This change does not alter the performance of the pressure and 
differential pressure transmitters and switches used in the plant 
protection systems. All sensors will still have response time 
verified by test before placing the sensors in operational service 
and after any maintenance that could affect response time. Changing 
the method of periodically verifying instrument response for certain 
sensors from time response testing to calibration will not create 
any new accident initiators or scenarios. Implementation of the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in margin of safety.
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for selected pressure and differential pressure 
sensors is modified to allow use of actual test data or engineering 
data. The method of verification combined with sensing line 
preventative maintenance provides assurance that the total system 
response is within the time limit defined in the safety analysis, 
since calibration tests will detect any degradation which might 
significantly affect pressure sensor response time and periodic 
sensing line flushing will minimize the potential for long-term 
buildup of contaminants which may impact sensing line response. 
Based on the above, it is concluded that the proposed license 
amendment request does not result in a reduction in margin with 
respect to plant 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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: James H. Miller, III, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Project Director: David B. Matthews.

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

    Date of amendment request: March 30, 1994 (TS 318).
    Description of amendment request: The proposed amendment consists 
of six parts, which the licensee has designated ``A'' through ``F'' in 
the analysis of no significant hazards considerations presented below. 
The six parts are:
    A. Mechanical pressure and differential pressure indicating 
switches in the Reactor Protection System (RPS) and Emergency Core 
Cooling System (ECCS) are being replaced with an Analog Transmitter/
Trip System (ATTS) for the Browns Ferry Nuclear Plant (BFN) Unit 3.
    B. The BFN Units 1 and 3 reactor vessel water level safety limit is 
being revised to reflect the analytical limit provided by General 
Electric and the Level 1 Low Reactor Vessel Water Level setpoint is 
being revised to provide a more conservative limit.
    C. For BFN Unit 2, RPS and ECCS instrument identifiers are being 
added or corrected to enhance useability of the Technical 
Specifications. These changes do not reflect a change in equipment, 
operation of the associated system, or the safety function of that 
system.
    D. For BFN Unit 2, Reactor High Water Level, Reactor Core Isolation 
Cooling (RCIC) and High Pressure Coolant Injection (HPCI) Turbine Steam 
Line High Flow, and Drywell Pressure instrumentation calibration 
frequencies and functional test descriptions are being revised to 
reflect current calculations and test methods. These changes do not 
reflect a change in equipment, operation of the associated system, or 
the safety function of that system.
    E. For BFN Units 1, 2, and 3, the differential pressure 
instrumentation, which actuates the pressure suppression chamber-
reactor building vacuum breakers, calibration frequency is being 
revised. In addition, tables that specify the minimum number of 
instrument channels per trip system, function, trip level setting, 
actions required, remarks, functional test, and instrument check are 
being added.
    F. Corrects the capitalization of terms used on the affected BFN 
Units 1, 2, and 3 TS pages in order to conform with the current TS 
Definitions section. This part also corrects spelling and 
capitalization of other words on the same pages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Part A: The Unit 3 modification, which involves the installation 
of an Analog Transmitter/Trip System (ATTS), replaces older devices 
with devices of more modern design that perform the same function.
    The initiation of control rod insertion to mitigate a design 
basis accident is contained in Chapter 14 of the BFN Final Safety 
Analysis Report (FSAR). There is no change in design bases, 
protective function (initiation of control rod insertion), 
redundancy, setpoints, or logic associated with the installation of 
the ATTS. The consequences of a failure of this equipment are no 
different than that of the original equipment. Since there is no 
change in any protective functions, nor the creation of any new 
operational conditions, the proposed amendment does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    Part B: The revision to the Units 1 and 3 reactor vessel water 
level safety limit and the Level 1 low reactor vessel water level 
setpoint do not reflect any change in plant equipment. The safety 
limit is being changed to reflect the actual analytical safety limit 
calculated by General Electric.
    The Level 1 low reactor vessel water level trip initiates the 
Core Spray and Low Pressure Coolant Injection Systems and isolates 
the Main Steam lines. These actions are taken to mitigate the 
consequences of a Loss of Coolant Accident. The change in the 
setpoint affects the timing of the operation of equipment necessary 
to mitigate the consequences of an accident. A setpoint calculation 
has been generated which ensures these safety functions are 
initiated in accordance with the design basis accident analysis 
presented in Chapter 14 of the Browns Ferry FSAR. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    Part C: The addition or correction of Unit 2 instrument 
identifiers is administrative in nature and does not reflect any 
modification to plant equipment. These administrative changes do not 
reflect any change to any precursor for the design basis events or 
operational transients analyzed in the Browns Ferry FSAR. There is 
also no change to any protective function or mitigating action for 
the design basis events or operational transients analyzed in the 
Browns Ferry FSAR. Therefore, the probability or consequences of an 
accident previously evaluated is not significantly increased.
    Part D: The change in Unit 2 reactor high water level and 
Reactor Core Isolation Cooling (RCIC) instrumentation functional 
test descriptions reflects the equipment currently installed and the 
functional tests currently being performed.
    The changes in calibration frequencies are being made to reflect 
current setpoint calculations. There are no modifications to plant 
equipment or changes in instrument setpoints associated with these 
changes. The calibration frequencies specified by the current 
setpoint calculations ensure that the associated safety functions 
are initiated in accordance with the design basis accident analysis 
presented in Chapter 14 of the Browns Ferry Final Safety Analysis 
Report (FSAR). Therefore, the probability or consequences of an 
accident previously evaluated is not significantly increased.
    Part E: The changes in Units 1, 2, and 3 calibration frequency 
for the differential pressure instrumentation, which actuates the 
pressure suppression chamber-reactor building vacuum breakers, is 
being made to reflect current setpoint calculations. The specified 
minimum number of instrument channels per trip system, function, 
trip level setting, actions required, remarks, functional test, and 
instrument check reflect current operational requirements. There are 
no modifications to plant equipment or changes in instrument 
setpoints associated with these changes. The calibration frequencies 
specified by the current setpoint calculations ensure that the 
associated safety functions are initiated in accordance with the 
design basis accident analysis presented in Chapter 14 of the Browns 
Ferry Final Safety Analysis Report (FSAR). Therefore, the 
probability or consequences of an accident previously evaluated is 
not significantly increased.
    Part F: The proposed correction of the capitalization of terms 
in order to conform with the current TS Definitions section is 
administrative in nature and does not reflect any modification to 
plant equipment. The correction of spelling and capitalization of 
other words on the same pages is also administrative in nature and 
does not reflect any modification to plant equipment. These 
administrative changes do not reflect any change to any precursor 
for the design basis events or operational transients analyzed in 
the Browns Ferry FSAR. There is also no change to any protective 
function or mitigating action for the design basis events or 
operational transients analyzed in the Browns Ferry FSAR. Therefore, 
the probability or consequences of an accident previously evaluated 
is not significantly increased.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Part A: The installation of the ATTS replaces older devices with 
devices of more modern design that perform the same function. No new 
control functions are added. No credible equipment failure modes or 
single failure are introduced which could result in the inability of 
redundant safety components or systems to perform their safety 
functions in accordance with the design basis accident analysis 
presented in Chapter 14 of the Browns Ferry FSAR. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Part B: The revision to the Units 1 and 3 reactor vessel water 
level safety limit and the Level 1 low reactor vessel water level 
setpoint do not reflect any change in plant equipment. The safety 
limit is being changed to reflect the actual analytical safety limit 
calculated by General Electric.
    The change in the Level 1 low reactor vessel water level 
setpoint affects the timing of the operation of equipment necessary 
to mitigate the consequences of an accident. No new failure modes or 
system interactions are introduced. The same protection functions 
will still occur at the Level 1 low reactor water level setpoint. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Part C: The addition or correction of Unit 2 instrument 
identifiers is administrative in nature and does not reflect any 
modification to plant equipment. The correction of instrument 
identifiers does not require new system alignments, modifications, 
or changes in operating procedures. Therefore, no new external 
threats, system interactions, release pathways, equipment failure 
modes, or types of operator errors are created. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Part D: The change in Unit 2 reactor high water level and RCIC 
instrumentation functional test descriptions reflects the equipment 
currently installed and the functional tests currently being 
performed. The changes in calibration frequencies are being made to 
reflect current setpoint calculations. There are no modifications to 
plant equipment or changes in instrument setpoints associated with 
these changes. No new failure modes or system interactions are 
introduced. The same protection functions will still occur at the 
same setpoints. Therefore, the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Part E: The changes in Units 1, 2, and 3 calibration frequency 
for the differential pressure instrumentation, which actuates the 
pressure suppression chamber-reactor building vacuum breakers, is 
being made to reflect current setpoint calculations. The specified 
minimum number of instrument channels per trip system, function, 
trip level setting, actions required, remarks, functional test, and 
instrument check reflect current operational requirements. There are 
no modifications to plant equipment or changes in instrument 
setpoints associated with these changes. No new failure modes or 
system interactions are introduced. The same protection functions 
will still occur at the same setpoints. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Part F: The proposed correction of the capitalization of terms 
in order to conform with the current TS Definitions section is 
administrative in nature and does not reflect any modification to 
plant equipment. The correction of spelling and capitalization of 
other words on the same pages is also administrative in nature and 
does not reflect any modification to plant equipment. The correction 
of spelling and capitalization does not require new system 
alignments, modifications, or changes in operating procedures. 
Therefore, no new external threats, system interactions, release 
pathways, equipment failure modes, or types of operator errors are 
created. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Part A: The installation of the ATTS replaces older devices with 
devices of more modern design that perform the same function. The 
replacement equipment will improve reliability, accuracy and 
response times. There are no changes in the systems' design basis, 
protective function, or logic arrangement. Instrument setpoints and 
calibration frequencies are supported by Unit 3 specific 
calculations. Therefore, the proposed amendment does not involve a 
significant reduction in the margin of safety.
    Part B: The revision to the Units 1 and 3 reactor vessel water 
level safety limit and the Level 1 low reactor vessel water level 
setpoint do not reflect any change in plant equipment. The safety 
limit is being changed to reflect the actual analytical safety limit 
calculated by General Electric.
    The change in the Level 1 low reactor vessel water level 
setpoint is supported by a Unit 3 specific setpoint calculation that 
has been performed in accordance with the methodology endorsed by 
Regulatory Guide 1.105, Instrument Setpoints for Safety Related 
Systems. Therefore, the proposed amendment does not involve a 
significant reduction in the margin of safety.
    Part C: The addition or correction of Unit 2 instrument 
identifiers is administrative in nature and does not reflect any 
modification to plant equipment. Therefore, the proposed amendment 
does not involve a significant reduction in the margin of safety.
    Part D: The change in Unit 2 reactor high water level and RCIC 
instrumentation functional test descriptions reflects the equipment 
currently installed and the functional tests currently being 
performed. The changes in calibration frequencies are being made to 
reflect Unit 2 specific setpoint calculations. These calculations 
have been performed in accordance with the methodology endorsed by 
Regulatory Guide 1.105. There are no modifications to plant 
equipment or changes in instrument setpoints associated with these 
changes. Therefore, the proposed amendment does not involve a 
significant reduction in the margin of safety.
    Part E: The changes in Units 1, 2, and 3 calibration frequency 
for the differential pressure instrumentation, which actuates the 
pressure suppression chamber-reactor building vacuum breakers, is 
being made to reflect current setpoint calculations. The specified 
minimum number of instrument channels per trip system, function, 
trip level setting, actions required, remarks, functional test, and 
instrument check reflect current operational requirements. The 
setpoint calculations have been performed in accordance with the 
methodology endorsed by Regulatory Guide 1.105. There are no 
modifications to plant equipment or changes in instrument setpoints 
associated with these changes. Therefore, the proposed amendment 
does not involve a significant reduction in the margin of safety.
    Part F: The proposed correction of the capitalization of terms 
in order to conform with the current TS Definitions section is 
administrative in nature and does not reflect any modification to 
plant equipment. The correction of spelling and capitalization of 
other words on the same pages is also administrative in nature and 
does not reflect any modification to plant equipment. Therefore, the 
proposed amendment 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: Athens Public Library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: March 31, 1994 (TS 339).
    Description of amendment request: The proposed amendment consists 
of five parts, which the licensee has designated ``A'' through ``E'' in 
the discussion of no significant hazards considerations, below. These 
five parts are:
    A. For the Browns Ferry Nuclear Plant (BFN) Units 1 and 3, improve 
operating flexibility by expanding the allowable operating domain. This 
expansion is accomplished by revising the equations for the flow-biased 
Average Power Range Monitor (APRM) flux scram trip setting and the APRM 
rod block trip setting.
    B. Revises the Rod Block Monitor (RBM) limiting conditions for 
operation (LCOs) to require two RBM channels to be operable if the 
plant is operating with low thermal margins.
    C. Miscellaneous editorial changes to the BFN Units 1, 2, and 3 
technical specifications (TS).
    D. Revises the BFN Units 1, 2, and 3 TS to delete the specific 
value for the rated loop recirculation flow rate.
    E. Revises the BFN Units 1, 2, and 3 TS to relocate the specific 
equations for the APRM rod block and RBM upscale setpoint equations 
from the TS to the Core Operating Limits Report (COLR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Part A: The proposed change will permit expansion of the current 
allowable power/flow operating region to allow operation in the 
extended load line limit (ELLL) region. Operation of BFN Units 1 and 
3 in the ELLL region will not increase the probability of any 
accident previously evaluated since the Average Power Range Monitor 
(APRM) system and flow-biased scram setpoint are not identified as 
initiators of any design basis accidents or transients. 
Additionally, no credit is taken for the APRM flow biased scram in 
any accident or transient analyses. Therefore, the proposed change 
can not significantly increase the probability of an accident 
previously evaluated.
    TVA's analysis of operation in the ELLL region verified that the 
consequences of previously evaluated accidents are within the 
acceptance criteria of the licensing basis. Therefore, the proposed 
change does not involve an increase in the consequences of an 
accident previously evaluated.
    Part B: The proposed change does not increase challenges or 
create any new challenges to safety-related systems or equipment, or 
other equipment whose failure could cause an accident. The proposed 
change does not change the function of the rod block monitor (RBM) 
subsystem. The RBM subsystem will continue to block control rod 
withdrawal to ensure that fuel safety limits are protected. The 
revised RBM limiting conditions for operation and surveillance 
requirements provide increased assurance that the RBM will function 
to ensure that fuel safety limits are protected. Therefore, the 
proposed change does not involve an increase in the probability of 
an accident previously evaluated.
    The revised RBM operability and surveillance requirements 
provide increased assurance that the RBM will block control rod 
withdrawal to ensure that fuel safety limits are protected. 
Accordingly, operation of BFN Units 1 and 3 with the revised RBM 
upscale setpoint does not involve an increase in the consequences of 
an accident previously evaluated.
    Part C: The miscellaneous editorial changes do not affect any 
plant operations, equipment, or any safety-related activity. These 
changes increase the probability that the specifications will be 
correctly interpreted by adding clarifying information and/or 
correcting errors. Therefore, these editorial changes do not involve 
an increase in the probability or consequences of an accident 
previously evaluated.
    Part D: The proposed change will delete the specific value for 
the rated loop recirculation flow rate found in the limiting safety 
system settings. This flow rate is in the TS to provide additional 
information, and is not a TS requirement. The proposed change does 
not change the limiting safety system settings or alter the method 
for calculating the settings. The proposed change does not affect or 
change operation of the plant, plant equipment, or any safety-
related equipment. The proposed change does not change the APRM rod 
block or trip settings, the method or frequency of calibration of 
the APRM flow biased network, or any other operational features of 
the APRM system. The proposed change will only delete an incorrect 
flow rate from the TS. Therefore, the proposed change does not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    Part E: The proposed change will remove specific equations for 
the APRM rod block and RBM upscale trip setpoints from the TS and 
relocate them to the Core Operating Limits Report (COLR). Removing 
these equations from the TS does not affect or change the APRM and 
RBM subsystems or the functions of these systems. The proposed 
change does not affect or change operation of the plant, plant 
equipment, or any safety-related equipment. Accordingly, the 
proposed change does not involve an increase in the probability of 
an accident previously evaluated.
    Removing the specific rod block equations from the TS does not 
change the requirements to comply with the limits of these equations 
during plant operations, since the TS will reference the COLR as the 
source of the equations. The actions to be taken in the event of 
noncompliance with the COLR-specified equations will also remain 
unchanged. Both the APRM rod block and RBM subsystems will continue 
to block control rod withdrawal to prevent reactor power from 
increasing to excess levels and to ensure that applicable limits of 
the plant safety analysis are met. Additionally, in accordance with 
the requirements of TS 6.9.1.7, these equations will continue to be 
developed using NRC-approved methodologies and will continue to 
ensure that applicable safety limits are protected. Therefore, the 
proposed change does not involve an increase in the consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Part A: Operation of BFN Unit 1 and 3 in the ELLL region does 
not create any new failure mode or sequence of events that can lead 
to an accident of a different type than any previously evaluated. 
Operation in the ELLL region does not increase challenges or create 
any new challenges to safety-related systems or equipment, or other 
equipment whose failure could cause an accident. Changing the 
equation for the flow-biased APRM scram trip setpoint does not 
change the function of the APRM subsystem. The APRM scram trip 
setpoint will continue to initiate a scram to ensure that the fuel 
safety limit is not exceeded. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Part B: The proposed change to the RBM operability and 
surveillance requirements does not create any new failure mode or 
sequence of events that can lead to an accident of a different type 
than any previously evaluated. The proposed change does not increase 
challenges or create any new challenges to safety-related systems or 
equipment, or other equipment whose failure could cause an accident. 
The proposed change does not change the function of the RBM 
subsystem. The RBM subsystem will continue to block control rod 
withdrawal to ensure that fuel safety limits are protected. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Part C: The miscellaneous editorial changes do not affect any 
plant operations, equipment, or any safety-related activity. These 
changes increase the probability that the specifications will be 
correctly interpreted by adding clarifying information and/or 
correcting errors. Therefore, these editorial changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Part D: The proposed change will delete the specific value for 
the rated loop recirculation flow rate found in a factor used to 
calculate limiting safety system settings for the APRM rod block and 
trip settings. The proposed change does not change the limiting 
safety system settings or alter the method for calculating the 
settings. The proposed change does not affect or change operation of 
the plant. The proposed change does not change the APRM rod block or 
trip settings, the method or frequency of calibration of the APRM 
flow biased network, or any other operational features of the APRM 
system. The proposed change will only delete an incorrect flow rate 
that is required to be calculated and maintained outside of the TS.
    Since there will be no change in plant operations, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Part E: Removal of the APRM rod block and RBM upscale setpoint 
equations does not change or affect any safety-related plant 
equipment or its functions; or any equipment, systems, or setpoints 
designed to prevent or mitigate accidents. Removing these rod block 
equations does not create any new challenges to safety-related 
systems or equipment, or other equipment whose failure could cause 
an accident; and does not change the function and manner of 
operation of the APRM or RBM subsystems. The APRM and RBM subsystems 
will continue to block control rod withdrawal to prevent reactor 
power from increasing to excess levels and to ensure that fuel 
safety limits are protected. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Part A: Operation of BFN Units 1 and 3 in the ELLL region does 
not affect the ability of the plant safety-related trips or 
equipment to perform their intended functions. Operation in the ELLL 
region will not cause any significant increase in offsite radiation 
doses resulting from any analyzed event. Although this change 
increases the APRM flow-biased scram equation, no credit is taken 
for this equation in the accident analyses. These analyses assume 
that transient events initiated from less than rated conditions are 
terminated by the fixed 120% flux scram or other safety-grade scram 
signals. These signals are not affected by the proposed change. 
Additionally, as noted above, TVA's analysis of operation in the 
ELLL region determined that the consequences of previously evaluated 
accidents remain within the acceptance criteria of the licensing 
basis. Therefore, this change does not involve a reduction in a 
margin of safety.
    Part B: The proposed change does not change the function of the 
RBM system. The RBM system will continue to block control rod 
withdrawal to ensure that fuel safety limits are protected. The 
proposed change does not affect plant operation, design, or any 
safety-related activity or equipment. The proposed change does not 
affect or change any margin of safety. The proposed change will 
actually increase the margin of safety by providing more 
conservative operability and surveillance requirements for the RBM 
subsystem. Therefore, the proposed change does not involve a 
reduction in a margin of safety.
    Part C: The miscellaneous editorial changes do not affect plant 
operation, design, or any safety-related activity or equipment. 
These changes increase the probability that the specifications will 
be correctly interpreted by adding clarifying information and/or 
correcting errors. Therefore, these changes do not involve a 
reduction in a margin of safety.
    Part D: The proposed change will delete the specific value for 
the rated loop recirculation flow rate found in the limiting safety 
system settings. The proposed change does not change the limiting 
safety system settings or alter the method for calculating the 
settings. The proposed change does not affect or change any margin 
of safety. The proposed change does not alter the APRM rod block or 
trip settings, nor does it change the combinations of power and flow 
conditions which could produce the APRM flow biased rod block and 
scram trips. Furthermore, the value for rated loop recirculation 
flow rate will continue to be contained in plant procedures which 
are controlled by the 10 CFR 50.59 process. Therefore, the proposed 
change does not involve a reduction in a margin of safety.
    Part E: The proposed change to remove the APRM rod block and RBM 
upscale setpoint equations does not change the equations or alter 
the method for calculating the equations. The proposed change does 
not change or affect any safety-related plant equipment or its 
functions; or any equipment, systems, or setpoints designed to 
prevent or mitigate accidents. Removing these rod block equations 
does not create any new challenges to safety-related systems or 
equipment, or other equipment whose failure could cause an accident; 
and does not change the function and manner of operation of the APRM 
or RBM subsystems. The requirements of TS 6.9.1.7 will continue to 
ensure that these equations are developed using NRC-approved 
methodology, and are consistent with applicable limits of the plant 
safety analysis. Therefore, the proposed change does not involve a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: August 9, 1994.
    Brief description of amendments: The proposed amendments would 
revise the technical specifications by eliminating the high negative 
neutron flux rate trip function.
    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 deletion of the High Negative Neutron Flux Rate trip does 
not adversely affect the probability of any accident. In fact, the 
deletion of this reactor trip is expected to reduce the probability 
of inadvertent reactor trips during surveillance testing.
    The only previously evaluated accidents whose consequences could 
be potentially affected by this change are the dropped rod events. 
Presently the High Negative Neutron Flux Rate trip function responds 
to these events by initiating a reactor trip. Analyses of these 
events, using currently licensed analysis methodologies have 
demonstrated that this trip function is not necessary. Although the 
scenarios for the various rod drop events without this trip function 
differ from the existing event scenarios, the unit will either be 
safely shutdown or return to an acceptable reactor power level, and 
as before, DNB does not occur. Therefore, the proposed changes would 
not have a significant effect on the 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 remove a feature from the Reactor Trip 
System that is a potential source of inadvertent or unnecessary 
reactor trips. While the changes delete an accident mitigation 
feature, they do not add new hardware to the units and do not change 
plant operations; consequently, no new failure modes are introduced. 
Therefore, the removal of this trip function cannot create the 
possibility of an [sic] 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.
    Because the changes only delete an unnecessary reactor trip and 
do not actually alter the plant, the proposed changes do not affect 
the failure values for any system or component. Accident analyses 
have shown that all relevant, existing event acceptance criteria 
have been satisfied without taking credit for the deleted reactor 
trip; therefore, the event acceptance criteria are not being 
revised. Because neither the failure values nor the acceptance 
criteria are affected, the proposed changes have no affect on the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: August 12, 1994.
    Brief description of amendments: The proposed amendments would 
revise the Administrative Controls section of the technical 
specifications to reflect changes to the licensee's organization, 
change the submittal due date for the Monthly Operating Report, and 
delete provisions that will be relocated to other controlled documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes are administrative in nature and reflect 
new organizational position titles, reassigned managerial 
responsibilities, a change in the required submittal due date of the 
Monthly Operating Report, and the relocation of selected Technical 
Specification Administrative Controls to licensee controlled 
documents. The proposed organizational changes include manager 
position title updates as follows: ``Group Vice President, Nuclear 
Engineering and Operations'' changed to ``Group Vice President, 
Nuclear Production''; ``Vice President, Nuclear Operation'' changed 
to ``Vice President of Nuclear Operations''; and ``Shift 
Supervisor'' changed to ``Shift Manager''. Manager responsibilities 
remain unchanged except the annual management directive concerning 
the Shift Supervisor's control room command function which is 
reassigned to the Group Vice President, Nuclear Production. 
Additionally, responsibilities of the Vice President, Nuclear 
Operations and Plant Manager are proposed for reassignment to either 
the Vice President of Nuclear Operations or the Plant Manager. In 
all cases, responsibilities will continue to be assigned to 
appropriately qualified individuals.
    The proposed change in the submittal due date of the Monthly 
Operating Report is to provide additional time to facilitate data 
compilation and report preparation, review and approval for two 
operating units. This change does not alter any data or information 
already reported.
    The proposed administrative control relocations include details 
of minimum shift operations crew staffing, method of onsite and 
offsite review and audit, selected reportable event actions, review 
of security and emergency plans and associated implementing 
procedures, requirements for record retention, and program elements 
for radiation protection, process control and radiological 
environmental monitoring. These administrative controls are 
addressed by other regulatory requirements and are relocated to 
other licensee documents (i.e., Final Safety Analysis Report, 
Emergency Plan, Security Plan, Offsite Dose Calculation Manual) 
which have adequate change control to ensure that intended plant 
design/safety functions will be maintained. No design basis 
accidents are affected by these proposed administrative changes as 
they do not impact nor affect accident analysis assumptions. 
Therefore, accident analyses assumptions are preserved and there is 
no change in the probability or consequences of any previously 
evaluated accident.
    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 to update organizational position titles 
and reassign responsibilities will not delete any responsibility/
function already designated in the Technical Specifications. All 
given management activities will continue to be performed by 
qualified individuals. The proposed change in the submittal due date 
of the Monthly Operating Report does not alter any data or 
information already reported. The administrative control relocations 
retain adequate regulatory basis to ensure that intended plant 
design/safety functions will be maintained. These changes are 
administrative in nature and do not affect the design or operation 
of any system, structure, or component in the plant. Accordingly, no 
new failure modes have been defined for any plant system or 
component important to safety, nor have any new initiating events 
been identified as a result of the proposed changes. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes involve the Administrative Controls Section 
of the CPSES Units 1 and 2 Technical Specifications and provide for 
updating organizational position titles, reassigning managerial 
responsibilities/functions, changing the required submittal due date 
for the Monthly Operating Report, and relocating administrative 
controls to other controlled licensee documents. These changes are 
administrative in nature and do not directly affect any protective 
boundaries nor impact the safety limits for the protective 
boundaries. The relocated requirements retain adequate regulatory 
basis for continued proper administrative review and plant 
configuration control to ensure that actions prescribed in plant 
operating procedures are maintained so as not to impact the plant's 
margin of safety. Therefore, there is no significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: August 4, 1994.
    Description of amendment request: This proposed amendment would 
reflect changes in the boron dilution accident analysis to address 
Boron Dilution Mitigation System (BDMS) time delays, the BDMS actuation 
setpoint uncertainty, and concerns regarding the applicability of the 
assumed inverse count rate ratio (ICRR) curve. The Technical 
Specifications include:
    a. Changing Bases page 2-8, note** of Table 3.3-1, and note 12 of 
Table 4.3-1 to reflect ``flux multiplication'' rather than ``flux 
doubling;''
    b. Revising note 9 of Table 4.3-1 to reflect the revised setpoint;
    c. Revising Bases page 3/4 4-1 to reflect new analysis assumptions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The actuation setpoint decrease and administrative controls to 
isolate dilution sources if no reactor coolant loop is in operation 
during Modes 3-5 reflect the new analysis assumptions. The 
initiating events are presented in FSAR Section 15.4.6. The proposed 
changes affect only the time required for BDMS to mitigate the event 
and do not affect the probability of any event initiators.
    Overall protection system performance will remain within the 
bounds of the accident analyses documented in FSAR Chapter 15, WCAP-
10961-P, and WCAP-11883 since no hardware changes are proposed.
    The BDMS will continue to function in a manner consistent with 
the above analysis assumptions and the plant design basis. As such, 
there will be no degradation in the performance of nor an increase 
in the number of challenges to equipment assumed to function during 
an accident situation.
    These Technical Specification revisions do not involve any 
hardware changes nor do they affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters or accident mitigation capabilities. Therefore, there 
will be no increase in the probability of any accident occurring due 
to the revised analysis.
    The results of this new analysis indicate that there is 
sufficient time for BDMS action to prevent a loss of plant shutdown 
margin. Since plant shutdown margin is not lost, the minimum DNBR 
remains well above the safety analysis limit values, no 
overpressurization occurs and, therefore, there are no fuel 
failures. The Technical Specification limits on shutdown margin in 
Modes 3-5 will be met. The conclusions of NRC Generic Letter 85-05 
and NSAC-183 remain valid (i.e., that gradual boron dilution events 
are self-limiting due to inherent reactivity feedback mechanisms). 
Given the above, there will be no increase in the consequences of 
any accident.
    (2) Create the possibility of a new or different kind of 
accident from any previously evaluated.
    As discussed above, there are no hardware changes associated 
with these Technical Specification revisions nor are there any 
changes in the method by which any safety-related plant system 
performs its safety function. The normal manner of plant operation 
is unaffected.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes. 
Therefore, the possibility of a new or different type of accident is 
not created.
    (3) Involve a significant reduction in a margin of safety.
    The results of the new analysis show that there is sufficient 
time for BDMS action to prevent a loss of plant shutdown margin. 
Since plant shutdown margin is not lost, the minimum DNBR remains 
well above the safety analysis limit values. The Technical 
Specification limits on shutdown margin in Modes 3-5 will be met. 
There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on DNBR limits, 
FQ, F-delta-H, LOCA PCT, peak local power density, or any other 
margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes to the Technical Specifications do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
or involve a significant reduction in a margin of safety. Therefore, 
it is concluded that the proposed changes meet the requirements of 
10CFR50.92(c) and does not involve a significant hazards 
consideration.

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

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

    Date of amendment request: August 4, 1994.
    Description of amendment request: The proposed amendment would 
replace Technical Specification (TS) 3/4.6.2.2, Spray Additive System, 
with a new TS 3/4.6.2.2 entitled Recirculation Fluid pH control (RFPC) 
System. The associated TS Surveillance Requirements and the Bases would 
be revised. In addition, the Bases for the Refueling Water Storage Tank 
(RWST) System will be revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Overall protection system performance will remain within the 
bounds of the accident analyses documented in FSAR Chapter 15, WCAP-
10961-P, and WCAP-11883.
    The accidents evaluated in the FSAR that could be affected by 
this proposed change are those involving the pressurization of the 
containment and associated flooding of the containment and 
recirculation of this fluid within the ECCS or the Containment Spray 
System (i.e., large break LOCA, main steam line break inside 
containment, and feedwater line break inside containment). The TSP-C 
will dissolve in the containment sump fluid resulting from these 
accidents raising the pH of the fluid, which would initially be 
greater than or equal to 4.0 but less than 7.0 during the injection 
phase of containment spray operation. The equilibrium spray pH 
during the recirculation phase resulting from this change will be 
greater than or equal to 7.1. The pH range for the spray will be 
bounded by the range of 4.0 to 11.0 in the current FSAR Section 
3.11(B) for the postulated spray solution environment. Since the 
resulting pH level will be closer to neutral using the TSP-C instead 
of NaOH, post-LOCA corrosion of containment components will not be 
increased. The results of the current, post-LOCA hydrogen generation 
calculation will remain bounding. There will not be an adverse 
radiation dose effect on any safety-related equipment. Thus, the 
potential for failures of the ECCS or safety-related equipment 
following a LOCA will not be increased as a result of the proposed 
change. The radiological consequences of changing from NaOH to TSP-C 
were reanalyzed using the current NRC methodology presented in 
Revision 2 of the Standard Review Plan (NUREG-0800) Section 6.5.2. 
This reanalysis indicates that the proposed change would result in 
reduced control room doses. Offsite doses would remain less than 
those currently reported in FSAR Table 15.6-8. The offsite and 
control room doses will continue to meet the requirements of 10 CFR 
100, 10 CFR 50 Appendix A GDC 19, SRP 15.6.5.II, and SRP 6.4.II. The 
dose reanalysis, combined with knowledge gained from recent studies 
on the behavior of iodine in the post-LOCA environment, demonstrates 
that the deletion of the Spray Additive System and replacement with 
a sump pH control system using TSP-C will not increase the reported 
radiological consequences of a postulated LOCA. The proposed new pH 
control system will provide satisfactory retention of iodine in the 
sump water, as well as provide adequate pH control to minimize the 
potential of chloride-induced stress corrosion cracking of 
austenitic stainless steel components.
    The Containment Spray System will continue to function in a 
manner consistent with the plant design basis. There will be no 
degradation in the performance of nor an increase in the number of 
challenges to equipment assumed to function during an accident 
situation.
    These Technical Specification revisions do not affect the 
probability of any event initiators. There will be no change to 
normal plant operating parameters, ESF actuation setpoints, or 
accident mitigation capabilities. Therefore, these changes will 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.
    The new Recirculation Fluid pH Control System is a passive 
system, i.e., no operator or automatic action is required to actuate 
the system. There are no active components being added whose failure 
could prevent the new system from functioning. The only new 
components being added are the TSP-C storage baskets. Seismic 
requirements have been included in the design to ensure the 
structural integrity of the baskets will be maintained during a 
seismic event.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes. 
The use of dry sodium phosphates is allowed for adjustment of the 
post-LOCA sump solution pH as discussed in SRP 6.1.1. Trisodium 
phosphate has a dissolution rate of 0.7 lbm/ft2-min in water at 
160 deg.F (given in WCAP-12477, based on trisodium phosphate in the 
form of a solid block with no agitation of the solution). The 
quantity of trisodium phosphate chosen will provide a minimum 
equilibrium sump pH of 7.1 following dissolution and mixing. No new 
equipment performance burdens are imposed' however, there is the 
potential for an unlikely, but possible, event in which an initially 
concentrated solution of TSP-C occupies the stagnant volume of an 
inoperable sump. This situation would not last for long since, as 
the recirculated sump fluid is cooled in the RHR heat exchangers, 
sufficient buoyancy-driven circulation within containment will 
result to displace the stagnant solution and eventually yield a 
uniform, equilibrium solution. Therefore, the possibility of a new 
or different type of accident is not created.
    (3) Involve a significant reduction in a margin of safety.
    The radiological analysis performed for this proposed change, as 
discussed above, shows that there would be no impact on the doses 
reported in FSAR Table 15.6-8.
    There will be no change to the DNBR Correlation Limit, the 
design DNBR limits, or the safety analysis DNBR limits discussed in 
Bases Section 2.1.1.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on DNBR limits, 
FQ, F-delta-H, LOCA PCT, peak local power density, or any other 
margin of safety.
    Based upon the preceding information, it has been determined 
that the proposed changes to the Technical Specifications do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
or involve a significant reduction in a margin of safety. Therefore, 
it is concluded that the proposed changes meet the requirements of 
10 CFR 50.92(c) and 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: January 6, 1994.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to (1) modify a table notation 
which would allow reactor protection channels to be placed in an 
inoperable status for a defined period of time to allow conduct of 
surveillance testing without requiring entry into the associated 
limiting condition for operation (LCO) action requirements, and (2) 
delete the channel check requirements for the ``Reactor Steam Dome--
High'' TS surveillance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    Regarding the proposed modifications to the Instrumentation Section 
table notations:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed revised notations to the TS Instrumentation Section 
tables clarify the current intent of the table notations by allowing 
the conduct of surveillance testing without declaring the affected 
channels inoperable for a limited period of time. In addition, the 
proposed changes make the wording of the table notations consistent 
with the Improved Technical Specifications (ITS). The proposed changes 
do not affect the time that the current notations allow a channel to be 
placed in an inoperable status, and thus do not affect the probability 
of an accident previously evaluated. In addition, the proposed changes 
do not affect the current requirement that redundant instrumentation be 
operable to accomplish the required function for the channel placed in 
an inoperable status. Thus, the proposed changes do not affect the 
consequences of an accident previously evaluated.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not affect the operation or availability of 
equipment from the current TS, and therefore do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The margin of safety affected by the proposed changes is the time 
that instrumentation required for plant protection is unavailable to 
perform its required function(s). The proposed changes do not affect 
the time that the instrumentation is allowed to be in an inoperable 
status during surveillance testing, and thus does not affect the 
availability of required instrumentation. In addition, the proposed 
changes do not affect the current requirement that redundant 
instrumentation remains operable, ensuring the required safety function 
remains operable during surveillance testing. Thus, the margin of 
safety is not affected by the proposed changes.
    Regarding the proposed removal of the channel check requirement for 
the ``Reactor Vessel Steam Dome Pressure--High'' instrumentation:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed removal of this channel check requirement would not 
affect the availability of the trip function associated with the 
``Reactor Vessel Steam Dome Pressure--High'' instrument, since the 
status of the channels is monitored by existing annunciators. This 
ensures the instruments remain capable of performing their intended 
function.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical modification to the 
facility or its equipment, nor does it result in a change in the 
reliability of the equipment. The proposed change does not, therefore, 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The margin of safety associated with not performing channel checks 
on the ``Reactor Vessel Steam Dome Pressure--High'' instrument is 
defined by the plant response to accidents that depend on this 
instrument, which in turn is dependent on the likelihood of 
availability of the trip function associated with this instrument. The 
proposed change does not affect the availability of this instrument, 
since the operators have continuous indication of the instrument's 
availability. Thus, the amendment does not affect 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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Theodore R. Quay.

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

    Date of amendment request: April 11, 1994.
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
3.1.f, ``Minimum Conditions for Criticality,'' in preparation for the 
licensee's extension of the fuel cycle from 12 to 18 months. The 
proposed amendment would revise TS 3.1.f to specify that the moderator 
temperature coefficient (MTC) shall be no greater than 5.0 pcm/ deg.F 
when at or below 60% rated thermal power and shall be zero or negative 
when above 60% rated thermal power. The proposed amendment also 
incorporates required actions to be implemented if the MTC 
specification is not met.
    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 TS 3.1.f

    This change is being proposed in accordance with the provisions 
of 10 CFR 50.92 to show that no significant hazards exist. The 
proposed change will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Moderator temperature coefficient (MTC) is a physical 
characteristic of the reactor core which affects the reactor's 
response to transients. The MTC is not an accident initiator, 
therefore changing the allowed MTC to 5.0 pcm/ deg.F at or below 60% 
power, does not increase the probability of an accident previously 
evaluated.
    To verify that an increased MTC would not increase the 
consequences of a previously evaluated accident, Wisconsin Public 
Service Corporation (WPSC) reanalyzed the bounding MTC-related 
events found in Kewaunee's Updated Safety Analysis Report (USAR). 
Those events are:
    1. Uncontrolled Rod Withdrawal at Intermediate Power (USAR 
Section 14.1.2)
    2. Rod Ejection at Hot Zero Power, Beginning of Cycle (USAR 
Section 14.2.6)
    3. Loss of External Electrical Load, Beginning of Cycle (USAR 
Section 14.1.9)
    4. Uncontrolled Boron Dilution (USAR Section 14.1.4)
    5. Loss of Reactor Coolant Flow, Locked Rotor (USAR Section 
14.1.8)
    6. Loss of Reactor Coolant Flow, Both Pumps Trip (USAR Section 
14.1.8)
    Events numbered 1, 3, 4, 5, and 6 above were reanalyzed to 
verify that the departure from nuclear boiling ratio (DNBR) would 
not fall below the acceptance criterion of 1.300. In all five cases, 
an assumed MTC value of 5.0 pcm/ deg.F at or below 60% power 
increased the calculated DNBR value from the current USAR analyses. 
Therefore the existing analyses associated with DNBR remain 
bounding.
    Event number 2 above was reanalyzed to ensure Kewaunee's peak 
clad temperature (PCT) would not exceed the required acceptance 
criterion of 2700 deg.F. When this accident was reanalyzed using the 
positive MTC value of 5.0 pcm/ deg.F, the calculated PCT exceeded 
2700 deg.F. Therefore it was necessary to decrease the allowable hot 
channel factor (FQ) from 9.0 to 8.2. This change in allowable FQ 
reduces a bounding value for future Kewaunee core designs and does 
not have an adverse effect on plant safety limits or settings. This 
FQ limit reduction for the accident of Rod Ejection at Hot Zero 
Power, Beginning of Cycle, will be incorporated into the WPSC Reload 
Safety Methodology and Kewaunee's USAR. The results of the analysis 
using a FQ of 8.2 and a MTC value of 5.0 pcm/ deg.F indicate a lower 
PCT than the previously bounding case.
    In addition, WPSC reviewed the effects of this change on 
Kewaunee's anticipated transient without scram (ATWS) analysis. This 
review demonstrated that the ATWS analysis was found to include 
assumptions bounding Kewaunee's proposed MTC limits.
    The reanalyses of the six most limiting MTC-related transients 
and review of the ATWS event demonstrate that this proposed change 
does not involve an increase in the potential consequences of any 
accidents previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The MTC is a physical characteristic of the reactor core and is 
not an accident initiator. A changed MTC changes the reactor's 
response to the postulated transients in chapter 14 of Kewaunee's 
USAR. Therefore an MTC of 5.0 pcm/ deg.F at or below 60% rated 
thermal power does not create the possibility of a new or different 
type of accident from any accident previously evaluated.
    (3) Involve a significant reduction in the margin of safety.
    The six most limiting MTC-related transients in chapter 14 of 
the KNPP USAR were reanalyzed. Five of the six were reanalyzed to 
verify the DNBR would not go below the DNBR acceptance criterion of 
1.300. The results of these five DNBR-related analyses demonstrate 
that the reanalyzed DNBR value is greater than the DNBR from the 
currently bounding safety analyses. The sixth transient was 
reanalyzed to verify the PCT limit of 2700 deg.F would not be 
exceeded. Reanalysis of this sixth accident, Rod Ejection at Hot 
Zero Power and Beginning of Cycle, indicated a need to lower the 
allowable hot channel factor (FQ). Reanalysis incorporating this 
more restrictive FQ value resulted in a calculated PCT value of 
2504 deg.F, which is less than the previous PCT value of 2585 deg.F. 
The new FQ limit will be incorporated into the next USAR revision, 
which will be submitted in accordance with 10 CFR 50.71.
    In addition, WPSC reviewed the effects of this change on 
Kewaunee's anticipated transient without scram (ATWS) analysis. This 
review demonstrated that the ATWS analysis was found to include 
assumptions bounding Kewaunee's proposed MTC limits.
    The reanalyses of the six most bounding transients and review of 
the ATWS event demonstrate that restricting the MTC value to 5.0 
pcm/ deg.F at or below 60% power does not reduce the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: May 20, 1994.
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
3.4, ``Steam and Power Conversion System,'' by modifying and clarifying 
the operability requirements for the main steam safety valves (MSSVs), 
auxiliary feedwater system (AFW), and the condensate storage tank 
system. The proposed amendment would eliminate inconsistencies within 
Technical Specification Section 3.4 and would provide the basis for 
operation of the AFW system below 15% reactor power.
    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 MSSV's 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 at Power, Uncontrolled Rod 
Cluster Control Assembly Withdrawal From a Subcritical Condition, 
Steam Generator Tube Rupture, and Anticipated Transients without 
Scram.
    The initial conditions in the analysis of a Loss of External 
Electrical Load Accident, Loss of Normal Feedwater Accident, 
Uncontrolled Rod Withdrawal at Power Accident, Uncontrolled Rod 
Cluster Control Assembly Withdrawal from a Subcritical Condition 
Accident and an Anticipated Transients without Scram Accident assume 
critical conditions. 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.
    Two operable MSSVs are capable of relieving the maximum steam 
generated during a Reactor Coolant Pump Locked Rotor Accident or a 
Steam Generator Tube Rupture from a subcritical condition. The 
maximum required steam relief rate resulting from a Steam Generator 
Tube Rupture is 270,000 lbm/hr, well below the relief capacity of 
1,532,076 lbm/hr of two MSSVs on a single steam generator. Also, 
this condition results in no increase in the radiological dose to 
the public. Two MSSVs on one steam generator are capable of 
relieving the steam generated corresponding to approximately 20% 
reactor power. This relief capacity is well above the maximum steam 
generation rate resulting from a Reactor Coolant Pump Locked Rotor 
Accident from a subcritical condition.
    In all cases, the relieving capacity of the MSSVs exceeds the 
maximum steam generation rate, 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 3.4.6.2.A.
--The addition of a seven (7) day Limiting Condition for Operation 
(LCO) for one steam supply to the turbine driven auxiliary feedwater 
pump.
--An addition to allow the AFW 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.

    An inconsistency currently exists between current TS 3.4.a.1.A.1 
and current TS 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 removing both AFW supply headers from 
service for 48 hours.
    Proposed TS 3.4.b.2.A 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 affect on the health and 
safety of the public.
    Proposed TS 3.4.b.5 permits 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 implemented 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.5 to plant start-ups, 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%. 
All other acceptance criteria of 10 CFR 50.46 were [also] 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.
    For these reasons, this change will have no adverse affect 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 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.
    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%. All other acceptance criteria of 10 CFR 50.46 were [also] 
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.
    For these reasons, these changes will not adversely effect the 
health and safety of the public or involve a significant reduction 
in 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 changes 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 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon.

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.

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

    Date of application for amendment: August 18, 1994.
    Brief description of amendment request: The proposed amendment 
would add the analytical method supplement entitled ``Calculative 
Methods for the CE Large Break LOCA Evaluation Model for the Analysis 
of CE and W Designed NSSS,'' CENPD-132, Supplement 3-P-A, dated June 
1985, to the list of analytical methods in TS Section 6.9.1.10 used to 
determine the PVNGS core operating limits. Additionally, further 
administrative changes are proposed to the list of analytical methods.
    Date of individual notice in Federal Register: September 6, 1994 
(59 FR 46069).
    Expiration date of individual notice: October 6, 1994.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of application for amendment: August 29, 1994.
    Brief description of amendment request: The proposed amendment 
would revise the combined Technical Specifications (TS) for the Diablo 
Canyon Power Plant (DCPP) Unit Nos. 1 and 2 to specify an alternate 
method of determing water and sediment content for new diesel fuel oil 
as specified in TS 3/4.8.1.1, ``A.C. Sources - Operating.''
    Date of individual notice in Federal Register: September 8, 1994 
(59 FR 46453).
    Expiration date of individual notice: October 7, 1994.
    Local Public Document Room location: California Polytechnical State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Notice of Issuance of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendments: January 4, 1994.
    Brief description of amendments: The amendments would implement 
recommended changes from Generic Letter (GL) 93-05, ``Line-Item 
Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.'' Specifically, the 
amendments implement changes to Technical Specifications corresponding 
to the following GL 93-05 line numbers: 4.1.2, 5.8, 5.14, 6.1, 7.5, 
8.1, 9.1, and 12. Proposed changes to line number 14 were reviewed and 
deemed unnecessary because the changes are already in the Technical 
Specifications.
    Date of issuance: September 2, 1994.
    Effective date: September 2, 1994.
    Amendment Nos.: 79, 66, and 51.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 22, 1994 (59 FR 
37513). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 2, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of application for amendments: October 27, 1993.
    Brief description of amendments: The amendments would revise TS 
6.9.1.8 to change the frequency for submitting the Radioactive Effluent 
Release Report from semiannual to annual, as allowed by the revised 
Title 10, Code of Federal Regulation, section 50.36a requirements. In 
addition, the references to this report would be changed from 
semiannual to annual to ensure consistency throughout the TS.
    Date of issuance: September 12, 1994.
    Effective date: September 12, 1994.
    Amendment No.: 80, 67, 52.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
Amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2861). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 12, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of application for amendments: June 8, 1994.
    Brief description of amendments: The amendments revise the Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Technical Specification 
4.8.1.1.1.b to extend the alternate 69 kV offsite power circuit 
surveillance frequency to accommodate the 24-month fuel cycles 
currently in use at Calvert Cliffs. As requested in Generic Letter 91-
04, ``Changes In Technical Specification Surveillance Intervals To 
Accommodate a 24-month Fuel Cycle,'' the licensee provided an 
evaluation in support of the change which concludes the effect on 
safety is small and does not invalidate any assumption in the plant 
licensing basis.
    Date of issuance: September 6, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 195 and 172.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37063). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: June 8, 1994.
    Brief description of amendments: The amendments revise the Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specification 
4.6.2.2.b to extend the Containment Air Cooling System surveillance 
frequency to accommodate the 24-month fuel cycle currently in use at 
Calvert Cliffs.
    Date of issuance: September 6, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 196 and 173.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37063). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: October 14, 1993.
    Brief description of amendment: The amendment revises TS 4.2.2 to 
remove the schedule for withdrawal of reactor vessel material 
surveillance specimens.
    Date of issuance: September 6, 1994.
    Effective date: September 6, 1994.
    Amendment No. 150.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59745). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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

    Date of application for amendment: March 3, 1993, as supplemented 
on October 22, 1993, and August 12, 1994.
    Brief description of amendment: The amendment will (1) add ANF-88-
133(P)(A), ``Qualification of Advanced Nuclear Fuels' PWR Design 
Methodology for Rod Burnups of 62 Gwd/MTU,'' to the approved 
methodologies list of Section 6.9.3.3.b, (2) clarify the wording in 
Sections 3.10.2.1 and 3.10.2.2.2 by describing more precisely how 
measurement uncertainty and engineering factors are considered, (3) 
correct the licensee's typographical error in Section 3.10.2.2, and (4) 
correct a reference to Section 6.9.3.3.b on page 3.10-16a of the TS 
basis.
    Date of issuance: September 16, 1994.
    Effective date: September 16, 1994.
    Amendment No. 151.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 14, 1993 (58 FR 
19472). The letter dated October 22, 1993, and August 12, 1994, 
provided clarifying information that did not affect the published 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 16, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of application for amendments: January 19, 1994.
    Brief description of amendments: The amendments increase the 
minimum reactor coolant system temperature required for criticality 
from 500 degrees Fahrenheit to 530 degrees Fahrenheit and delete other 
requirements and bases discussions that no longer apply to the 
facility.
    Date of issuance: September 8, 1994.
    Effective date: September 8, 1994.
    Amendment Nos.: 156 and 144.
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37065). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of application for amendments: The original submittal was 
incorrectly dated March 31, 1993. The Federal Register notice reflected 
the correct date of March 31, 1994. The licensee's submittal dated 
April 15, 1994, superseded the first submittal in its entirety. The 
submittal was further supplemented on July 26, 1994.
    Brief description of amendments: The amendments change the 
Engineered Safeguards Features Actuation System (ESFAS) Technical 
Specifications (TS) sections to provide allowed outage times for 
Automatic Actuation Channel surveillance testing and restoration time 
for an inoperable ESFAS Automatic Actuation Channel.
    Date of issuance: September 8, 1994.
    Effective date: September 8, 1994.
    Amendment Nos.: 157 and 145.
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24748) The April 15 and July 26, 1994, submittals 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 September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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: July 13, 1994.
    Brief description of amendments: The amendments revise Technical 
Specification 4.6.2 by changing the testing interval for the air or 
smoke flow test through each containment spray header from 5 to 10 
years, and remove an obsolete footnote.
    Date of issuance: September 6, 1994.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 147 and 129.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39583) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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: July 13, 1994.
    Brief description of amendments: The amendments revise Technical 
Specification 4.6.2 by changing the testing interval for the air or 
smoke flow test through each containment spray header from 5 to 10 
years.
    Date of issuance: September 6, 1994.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 123 and 117.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39583) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of application for amendments: June 9, 1994.
    Brief description of amendments: These amendments revise the 
Appendix A TSs relating to new fuel oil for emergency diesel 
generators. The amendments replace the qualitative examination of new 
fuel oil for water/sediment and particulate contamination with a 
quantitative examination for these properties.
    Date of issuance: September 8, 1994.
    Effective date: As of date of issuance to be implemented within 60 
days of issuance.
    Amendment Nos.: 182 and 63.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34662) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

    Date of amendment request: April 14, 1994.
    Brief description of amendments: The amendments revise Technical 
Specification 5.3.1, ``Fuel Assemblies,'' to permit fuel assembly 
reconstitution to restore the usefulness of fuel assemblies containing 
damaged or leaking fuel rods.
    Date of issuance: September 2, 1994.
    Effective date: September 2, 1994.
    Amendment Nos.: Unit 1--Amendment No. 65; Unit 2--Amendment No. 54.
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29628) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 2, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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

    Date of application for amendment: April 18, 1994.
    Brief description of amendment: License Amendment No. 81, issued on 
July 15, 1993, changed the numbering of Surveillance Requirement 
4.1.3.2, ``Control Rod Maximum Scram Insertion Times,'' but failed to 
renumber multiple references made to this surveillance. This amendment 
renumbers those references previously overlooked.
    Date of issuance: September 2, 1994.
    Effective date: Date of issuance.
    Amendment No.: 92.
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24749) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 2, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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

    Date of application for amendment: August 5, 1994.
    Brief description of amendment: The amendment modifies the Clinton 
Power Station Technical Specification Table 4.8.1.1.2-1, ``Diesel 
Generator Test Schedule,'' to exclude three valid failures of the 
Division 1 diesel generator from contributing towards an accelerated 
test schedule.
    Date of issuance: September 2, 1994.
    Effective date: September 2, 1994.
    Amendment No.: 93.
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (59 FR 42080 dated August 16, 1994). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice published August 16, 1994, also provided for 
an opportunity to request a hearing by September 15, 1994, but 
indicated that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 2, 1994.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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

    Date of application for amendments: November 11, 1992, as 
supplemented December 17, 1993, and March 29, 1994.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase the main steam safety valve 
setpoint tolerance for both units to plus or minus 3 percent and revise 
the emergency core cooling system section for Unit 2 to reflect a 
thermal power limitation resulting from the small break loss-of-coolant 
accident analysis when a safety injection cross-tie is closed.
    Date of issuance: September 9, 1994.
    Effective date: September 9, 1994.
    Amendment Nos.: 182 and 167.
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12262). The December 17, 1993, letter corrected errors in the Small 
Break Loss of Coolant Accident analysis, and the March 29, 1994, letter 
modified the requested amendment to include the requirement to reset 
the MSSV setpoints to plus 1 percent whenever they are found outside 
that range. These letters were responding to NRC staff concerns, and 
provided clarifying information which did not change the staff's 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 9, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: January 6, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Tables 3.2.7, 3.6.2a, 4.6.2a, 3.6.2b, and 4.6.2b to 
delete the automatic reactor scram and main steam line isolation 
functions of the Main Steam Line Radiation Monitor. Conforming changes 
are also being made to the Bases for these TSs and to the Bases for TS 
2.1.2. The changes are consistent with the NRC's Improved Standard 
Technical Specifications (NUREG-1433) and with NRC-approved Boiling 
Water Reactor Owners Group Licensing Topical Report NEDO-31400A, dated 
July 9, 1987 (safety evaluation dated May 15, 1991).
    Date of issuance: September 9, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 149.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7692). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 9, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: May 27, 1994.
    Brief description of amendment: The amendment temporarily allows 
the Operations Manager to not have a senior reactor operator license 
for Millstone 3, providing other conditions are met.
    Date of issuance: September 1, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 94.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34667).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 4, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Community-Technical College, Thames Valley Campus, 574 New London 
Turnpike, Norwich, Connecticut 06360.

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

    Date of application for amendment: August 10, 1993.
    Brief description of amendment: The amendment involves several 
changes to the radiological effluents portion of the Technical 
Specifications. One of the changes provides clarification of sampling 
and analysis requirements prior to venting or purging of the 
containment. A second portion of the change updates liquid effluent 
sampling and analysis requirements to reflect improvements in sample 
technology. The remaining changes are editorial in nature.
    Date of issuance: September 7, 1994.
    Effective date: September 7, 1994.
    Amendment No.: 90.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57855).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of application for amendment: March 28, 1994.
    Brief description of amendment: The proposed amendment would change 
the reactor vessel level setpoint for initiation of secondary 
containment isolation and standby gas treatment system from the ``low'' 
setpoint to the ``low-low'' setpoint.
    Date of issuance: September 9, 1994.
    Effective date: September 9, 1994.
    Amendment No.: 91.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24750).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 9, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Dates of application for amendment: January 3, 1994 and August 29, 
1994.
    Brief description of amendment: The proposed amendment would revise 
the requirements of Technical Specification (TS) 4.6.E.1.a, which 
currently specifies that a minimum of seven safety/relief valves shall 
be bench checked or replaced with a bench checked valve each refueling 
outage. The proposed amendment would change this specification to 
require the valves to be tested in accordance with the Section XI 
inservice testing requirements of the ASME Boiler and Pressure Vessel 
Code. The proposed change is consistent with the Improved Standard 
Technical Specifications, NUREG-1433.
    Date of issuance: September 15, 1994.
    Effective date: September 15, 1994.
    Amendment No.: 92.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10011).
    The August 29, 1994, letter submitted by the licensee provided 
historical test data related to the components involved in the TS 
amendment and did not affect the staff's initial no significant hazards 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 15, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Northern States Power Company, Docket No. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

    Date of application for amendments: January 29, 1993, as revised 
June 15, 1994.
    Brief description of amendments: The amendments reflect changes to 
core exit thermocouple action statements.
    Date of issuance: September 7, 1994.
    Effective date: September 7, 1994.
    Amendment Nos.: 112 and 105.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39594). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of application for amendments: May 7, 1993 (Reference LAR 93-
01).
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2 to modify TS 3/4.3.3.5, ``Remote Shutdown 
Instrumentation,'' to add remote shutdown control functions, to 
increase the Allowed Outage Time (AOT) for an inoperable remote 
shutdown Function (instrumentation and control) from 7 days to 30 days, 
to add an Action Statement that clarifies that separate entry is 
permitted for each function listed in Table 3.3-9, and to revise the 
associated TS Bases.
    Date of issuance: September 2, 1994.
    Effective date: September 2, 1994.
    Amendment Nos.: 94 and 93.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39057).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 2, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating 
Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: April 27, 1994.
    Brief description of amendments: These amendments delete TS 
Surveillance Requirement 4.4.1.1.1, which requires that the reactor 
recirculation pump discharge valve be demonstrated operable by 
performing a full-stroke test of the valve prior to reactor thermal 
power exceeding 25 percent of rated reactor thermal power.
    Date of issuance: September 7, 1994.
    Effective date: September 7, 1994.
    Amendment Nos. 76 and 37.
    Facility Operating License Nos. NPF-39 and NPF-85. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29628). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 7, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: June 30, 1994.
    Brief description of amendments: The amendments relocate selected 
recirculation and control rod block instrumentation setpoints from 
Technical Specifications (TS) Table 3.3.6-2, and Section 3/4.4.1 to the 
Core Operating Limits Report (COLR), thereby revising TS Section 
6.9.1.9 to document relocation of these items into the COLR.
    Date of issuance: September 8, 1994.
    Effective date: September 8, 1994.
    Amendment Nos. 77 and 38.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 2, 1994 (59 
FR 39595). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: March 28, 1994.
    Brief description of amendments: These amendments increase the 
surveillance interval for the functional test of the Reactor Protection 
System electrical power monitoring channels from every 6 months to each 
time the plant is in cold shutdown for a period of 24 hours, unless the 
test was performed in the previous six months.
    Date of issuance: September 8, 1994.
    Effective date: September 8, 1994
    Amendment Nos. 78 and 39.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27062). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: May 13, 1994, as supplemented 
by letter dated July 20, 1994.
    Brief description of amendments: These amendments revise TS 
Surveillance Requirement 4.8.1.1.2e.8, which requires diesel generator 
be retested within 5 minutes after completing a 24-hour endurance run, 
to incorporate requirements containeed in NUREG-1433.
    Date of issuance: September 12, 1994.
    Effective date: September 12, 1994.
    Amendment Nos. 79 and 40.
    Facility Operating License Nos. NPF-39 AND NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32234). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 12, 1994.
    No significant hazards consideration comments received; No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: May 10, 1994, and supplemented 
by letters dated August 19, 1994 and September 13, 1994; these 
supplements provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination.
    Brief description of amendments: These amendments revise the 
minimum low pressure cooling availability requirements in accordance 
with the staff's recommendations in NUREG-1433, ``Standard Technical 
Specifications General Electric Plants, BWR/4.''
    Date of issuance: September 16, 1994.
    Effective date: September 16, 1994.
    Amendments Nos.: 195 and 199.
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37079).The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 16, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendments: June 17, 1994 and supplement 
dated July 13, 1994.
    Brief description of amendments: The amendments changed the 
requirements to perform the Channel Functional Test of the Power 
Operated Relief Valve (PORV) position indication from quarterly to 
every 18 months and to exempt the PORV Block Valve position indication 
from performance of the Channel Functional Test if the PORV Block Valve 
is shut as required to isolate a PORV that cannot be manually cycled.
    Date of issuance: September 8, 1994.
    Effective date: As of date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment Nos. 155 and 136.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39596). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendments: April 12, 1994 as supplemented 
by letter dated July 15, 1994.
    Brief description of amendments: The amendments eliminate the 
required loss of offsite power plus an engineered safety feature 
actuation signal test following the 24-hour endurance test, revise the 
5-minute emergency diesel generator (EDG) hot restart test by 
separating it from the 24-hour endurance test and add a new 
surveillance requirement of a simple hot restart test following a 2-
hour loaded run of the EDG.
    Date of issuance: September 8, 1994.
    Effective date: Unit 1, effective as of its date of issuance and 
shall be implemented within 90 days of the date of issuance and; Unit 
2, effective as of its date of issuance and shall be implemented before 
restart from the eighth refueling outage currently scheduled to begin 
on October 15, 1994.
    Amendment Nos. 156 and 137.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27066). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendments: May 25, 1994.
    Brief description of amendments: These amendments would modify the 
table associated with 3/4 3.3.1, Radiation Monitoring Instrumentation; 
3/4 3.3.3.8, Radioactive Liquid Effluent Monitoring Instrumentation; 
and 3/4 3.3.9, Radioactive Gaseous Effluent Monitoring Instrumentation 
by relocating the specific radiation monitor numbers from the table to 
a cross reference in the Bases.
    Date of issuance: September 8, 1994.
    Effective date: September 8, 1994.
    Amendment Nos. 157 and 138.
    Facility Operating License Nos. DPR-70 and DPR-75. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34668).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendment: February 25, 1994.
    Brief description of amendment: The amendment changes the Technical 
Specifications to permit the submittal of the Radioactive Effluent 
Release Report on an annual rather than a semiannual basis; allow 
changes to the Offsite Dose Calculation Manual to be submitted in the 
Radioactive Effluent Release Report rather than in the monthly 
operating report; remove the title of Executive Vice President 
Operations from the TS; remove the list of audit frequencies from the 
TS and place them under Quality Systems management; change the title of 
Associate Manager, Health Physics to Radiation Protection Manager; 
remove references to specific letters; remove TS 6.4 on training; and 
correct various typographical errors and include items omitted in 
previous amendments.
    Date of issuance: September 6, 1994.
    Effective date: September 6, 1994.
    Amendment No.: 117.
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24752).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: February 16, 1994, as revised August 4, 
1994.
    Brief Description of amendments: The amendments change the fuel 
description in the TS to allow fuel reconstitution, to permit the use 
of ZIRLOTM clad fuel, and to remove unnecessary detailed 
descriptions of fuel and control rod assemblies.
    Date of issuance: September 8, 1994.
    Effective date: September 8, 1994.
    Amendment Nos.: 110 and 101.
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12366).
    The August 4, 1994, submittal reformatted Section 5.6.1 and 
restored and relocated to Section 5.6.1 the maximum enrichment limits 
deleted in the February 16, 1994 amendment request, but did not change 
the no significant hazards consideration as published in the Federal 
Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 8, 1994.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

    Date of application for amendments: September 28, 1993 (TS 93-10).
    Brief description of amendments: The amendments clarify the 
operability requirements for the fire suppression system flow path and 
incorporate additional guidance into an action statement requirement 
for spray and/or sprinkler systems inside containment.
    Date of issuance: September 13, 1994.
    Effective date: September 13, 1994.
    Amendment Nos.: 186 and 178.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59757). The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated September 13, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: January 27, 1993, clarified on 
April 20, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specification 6.3.1.2 to allow either the Health Physics Superintendent 
or the Health Physics, Operations Supervisor to be designated as 
Radiation Protection Manager (RPM).
    Date of issuance: September 6, 1994.
    Effective date: September 6, 1994.
    Amendment No.: 92.
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specification 6.3.1.2.
    Date of initial notice in Federal Register: April 14, 1993 (58 FR 
19490).
    The clarifying information did not change the initial proposed no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendments: March 30, 1994.
    Brief description of amendments: The amendments revise the NA-1&2 
TS High Head Safety Injection (HHSI) surveillance requirements by 
removing explicit numerical values and replacing them with broader non-
numerical requirements.
    Date of issuance: September 6, 1994.
    Effective date: September 6, 1994.
    Amendment Nos.: 188 and 169.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22017)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 6, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By October 28, 1994, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC 20555 and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: September 9, 1994, as supplemented by 
letters dated September 9 and September 16, 1994.
    Description of amendment request: The amendment revises the 
Technical Specifications (TS) by adding a footnote to Surveillance 
Requirement 4.8.1.1.2.d to allow operation of the emergency train AB 
bus with alternative operability testing in lieu of integrated testing 
required to confirm operability for loss of offsite power or in 
conjunction with safety injection. The footnote will remain in effect 
until shutdown following refuel 7.
    Date of issuance: September 16, 1994.
    Effective date: As of the date of issuance.
    Amendment No.: 98.
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
September 16, 1994.
    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.

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

    Date of application for amendment: August 18, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications to extend the 18-month battery charger and battery 
service test from their present required date of September 7, 1994, 
until just prior to core reload in the upcoming Unit 2 refueling 
outage.
    Date of issuance: September 6, 1994.
    Effective date: September 6, 1994.
    Amendment No.: 166.
    Facility Operating License No. DPR-74. Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
September 6, 1994.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: September 2, 1994.
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) 4.7.1.1.1.d.4, 4.7.1.1.1.d.5, 4.7.1.2.1.d.4, and 
4.7.1.2.1.d.5 to delete the requirement that the surveillance 
requirements for demonstrating service water system pump performance 
and for verifying the integrity of the deicing heaters be performed 
during shutdown. The amendment only deletes the requirement for 
performing these surveillances during shutdown; the revised TSs 
continue to require that these surveillances be performed at least once 
per 18 months. The revised TSs permit these surveillances to be 
performed during any operational condition.
    Date of issuance: Sepetember 13, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 57.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
September 13, 1994.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

    Dated at Rockville, Maryland, this 21st day of September 1994.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 94-23819 Filed 9-27-94; 8:45 am]
BILLING CODE 7590-01-P