[Federal Register Volume 59, Number 22 (Wednesday, February 2, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10202]


[[Page Unknown]]

[Federal Register: February 2, 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 January 7, 1994, through January 21, 1994. 
The last biweekly notice was published on January 19, 1994 (59 FR 
2859).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, And Opportunity For A Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies 
of written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By March 4, 1994, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendment request: November 4, 1993
    Description of amendment request: The proposed amendment revises 
Technical Specification 6.13.1 to provide use of alarming dosimeters in 
high radiation areas. This change includes newly revised 10 CFR Part 20 
requirement references and is consistent with NUREG-1413, Standard 
Technical Specifications - Westinghouse Plants, Specification 5.11.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. This change could involve a reduction in 
personnel radiation exposure by utilizing alarming dosimeters. This 
change does not involve any plant systems or components which could 
increase the probability of an accident. 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. This change could reduce the possibility of an accidental 
overexposure by alerting personnel when their maximum allowable 
exposure has been received. This change does not involve any plant 
systems or components. 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. Due to the nature of this 
proposed change, it is not related to any plant system. 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: Hartsville Memorial Library, 
Home and Fifth Avenues, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: S. Singh Bajwa

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

    Date of amendment request: August 27, 1993
    Description of amendment request: The proposed amendments would 
revise the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 
and 2, Technical Specifications (TS) regarding inspection requirements 
for pipe snubbers. The proposed changes implement Generic Letter (GL) 
90-09, and would affect the surveillance requirements of TS section 
4.7.8 and the bases for these requirements, section 3/4.7.8. 
Specifically, the amendment would change the existing inspection 
periods, visual inspection acceptance criteria, and functional test 
requirements. Additionally, there would be changes to the bases to 
include reference to GL 90-09, and other editorial changes would be 
made.
    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 amended surveillance requirement adds a table that addresses 
the maximum number of snubber failures that can be tolerated prior 
to reducing the inspection interval. This number is a function of 
the population size of a particular type of snubber. The revised 
requirement will allow the inspection intervals to be compatible 
with the 24 month fuel cycles, and provisions are included to extend 
the inspection interval up to 48 months. A provision is included to 
allow an evaluation to determine operability to justify continued 
operation with a snubber that is unacceptable.
    The purpose of the amendment request is to provide for 
alternative inspection intervals that take the size of the 
population of a snubber type into account. The proposed change 
provides the same confidence level and allows snubber inspection and 
corrective action to be performed during refueling outages. This 
allows the plant to avoid a mid-cycle outage due to a small number 
of snubber failures.
    The proposed change allows for a small percentage of snubbers in 
each category to fail the required visual examination without 
adjusting the inspection frequency. If a statistically significant 
percentage of snubbers fail, the visual examination inspection 
interval is reduced based on the percentage of failed snubbers.
    The proposed change has no direct or indirect impact on 
reactivity management activities.
    The change is not expected to have an impact on equipment 
failures. Any snubbers that fail to meet the visual examination 
acceptance criteria are either functionally tested in the as-found 
condition to verify continued acceptability, or an evaluation is 
performed to demonstrate the acceptability of continued operation 
with an unacceptable snubber. No new equipment is being introduced 
and no systems are operated in a configuration that has not been 
evaluated, so no new failure modes are introduced.
    The affected transients are the design basis earthquake and the 
spectrum of event initiating transients, with the capability of 
imposing significant dynamic loads or otherwise which impact the 
structural integrity of the Reactor Coolant System (RCS).
    The snubbers are installed to ensure the structural integrity of 
the RCS and required support systems. Their failure is passive in 
nature. The probability of a transient initiating event occurring is 
unrelated to the existence or condition of equipment that is 
designed to perform a mitigating function. The snubbers are 
installed to ensure an acceptable system response to a dynamic load, 
and their availability does not impact the frequency of occurrence 
of earthquakes or other transients resulting in significant dynamic 
loading.
    The revised testing provisions are designed to allow some 
flexibility while still maintaining a high probability that the 
installed snubbers will be capable of performing their intended 
function when required. The revised surveillances appropriately 
consider the size of the population of a particular type of snubber, 
and are sufficient to ensure the consequences of an accident will be 
unchanged when the revised requirements are implemented. By 
maintaining a statistically high level of confidence in the function 
of the plant's snubbers, the system response to transient initiating 
events will be as designed and thus, the off-site dose projected to 
occur of any affected transient will remain acceptably low.
    As previously stated, the revised surveillance provides a high 
confidence that the affected systems will remain intact and 
functional. Evaluation of the effects of operating with a degraded 
snubber is required to ensure that adequate margin exists to support 
continued plant operation. If this evaluation cannot adequately 
justify continued operation, the appropriate action statement will 
be applied. These provisions are sufficient to assure that the 
probability of an equipment malfunction will not increase.
    The consequences of equipment malfunction will not increase. 
Sufficient redundancy exists to accommodate the complete failure of 
one train of required equipment. The requisite electrical and 
physical separation are sufficient to ensure that the redundant 
train remains unaffected. This redundancy is adequate to ensure that 
the undetected failure of a snubber will not have a severe impact on 
overall system response to a transient.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The possibility of a new or different type of accident is not 
created by this change. No new or different equipment is being 
introduced, and no system will be operated in a different 
configuration without first having the effects of the new 
configuration evaluated. The new configuration would be system and/
or plant operation with a snubber installed that has failed its 
visual examination. The required evaluation must be sufficient to 
provide confidence that continued operation is acceptable; 
otherwise, the provisions of the action statement will be observed.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    No reduction in the margin of safety will occur as a result of 
this proposed change. As previously described, the controls in place 
will provide a high confidence the affected systems will continue to 
be functional. No significant increase in the rate of occurrence of 
undetected inoperable snubbers is expected to occur, and the 
allowable failures prior to applying an increased test frequency is 
still a small percentage of the total snubber population.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of amendment request: September 2, 1993, supplemented by 
letter dated January 7, 1994.
    Description of amendment request: The proposed amendment would 
revise the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 
and 2, Technical Specifications (TS) to allow replacement of the 125 
Volt DC Gould batteries with the new 125 Volt DC AT&T batteries and 
rephrase their design duty cycle. In addition, the proposed amendment 
would revise the batteries crosstie loading limitations and the 
crosstie breaker limitations. The associated Bases would also be 
revised to discuss the purpose for the crosstie limitations and to 
discuss design duty cycle requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The replacement AT&T battery has been selected to meet or exceed 
the design, functional, and operational requirements of those of the 
present Gould battery, including crosstie load limitations. The 
crosstie breaker limitation change to allow crosstie between two 
shutdown units is consistent with the Safety Evaluations issued with 
Technical Specification Amendment 5 for both Braidwood and Byron 
Stations. The remaining changes are administrative in nature or 
provide clarification to maintain consistency with other Technical 
Specifications and the Standard Technical Specifications.
    The overall design, function, and operation of the DC system and 
equipment has not been altered by these changes. The proposed 
changes do not affect any accident initiators or precursors and do 
not alter the design assumptions for the systems or components used 
to mitigate the consequences of an accident as analyzed in UFSAR 
Chapter 15. Therefore, there is no increase in the probability or 
consequences of an accident previously evaluated.
    B. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The replacement AT&T battery will provide the same functions as 
those of the present Gould battery and will be operated with the 
same types of operational controls. These limits will include 
battery float terminal voltage, individual cell voltage and 
electrolyte specific gravity, and crosstie loading. Crosstie 
conditions are allowed under the present Technical Specifications. 
The remaining changes are administrative in nature or provide 
clarification to maintain consistency with other Technical 
Specifications and the Standard Technical Specifications.
    The DC system and its equipment will continue to perform the 
same functions and be operated in the same fashion. The proposed 
change does not create any new or common failure modes. The proposed 
changes do not introduce any new accident initiators or precursors, 
or any new design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, the possibility 
of a new or different kind of accident from any accident previously 
evaluated has not been created.
    C. The proposed change does not involve a significant reduction 
in a margin of safety.
    The replacement AT&T battery will meet or exceed the design, 
functional, and qualification requirements of those of the present 
Gould batteries. The proposed Technical Specification limitations 
for the AT&T battery are derived from the same methodology and 
margins as those for the Gould battery. Increasing the crosstie 
loading limit takes advantage of the larger AT&T battery capacity 
with its increased design margin. The proposed change to the 
crosstie loading limit will continue to conservatively envelop the 
postulated design requirements. The remaining changes are 
administrative in nature or provide clarification to maintain 
consistency with other Technical Specifications and the Standard 
Technical Specifications.
    The inherent design conservatism of the DC system and its 
equipment has not been altered. The DC system and its equipment will 
continue to be operated with the same degree of conservatism. 
Therefore, there is no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of amendment request: October 28, 1993
    Description of amendment request: The proposed amendment would 
revise the Emergency Core Cooling System (ECCS) injection valve stroke 
times and ECCS response times for Motor-Operated Valve (MOV) 
modifications that increase injection valve stroke times.
    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 
because:
    The probability of an accident previously evaluated will not 
increase as a result of this change, because the only modification 
being performed is to the stroke times for the LPCS [Low Pressure 
Core Spray System], LPCI [Low Pressure Coolant Injection System], 
and HPCS [High Pressure Core Spray System] injection valves. 
Changing the opening or closing time of the injection valves for 
these ECCS systems does not cause any accident previously evaluated 
to occur. Therefore, modifying their stroke times will not increase 
the probability of occurrence for any accident previously evaluated.
    The consequences of a LOCA [Loss of Coolant Accident] are not 
significantly increased and do not exceed the previously accepted 
licensing criteria for this accident. GE [General Electric Company] 
has calculated the revised licensing basis PCT [Peak Centerline 
Temperature] for LaSalle Station to be 1260 deg.F, which is well 
below the 2200 deg.F criterion of 10 CFR 50.46 and Section 15.6.5 of 
NUREG-0800 (Standard Review Plan). The acceptance criteria for 
cladding oxidation, metal-water reaction (hydrogen generation), 
coolable geometry and long-term cooling also continue to be met with 
the increased valve stroke times.
    GE has performed sensitivity analysis justifying the continued 
applicability of previous analyses for Anticipated Transients 
Without Scram (ATWS), containment analyses, off-site dose (Main 
Steamline Break Outside Containment), and HPCS-related transients 
(Loss of Feedwater Flow). Other events are not affected because 
these systems are not assumed to function.
    2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because:
    The only modification is to increase the stroke time of the 
injection valves for LPCS, LPCI, and HPCS. This does not result in 
any changed component interactions, other than to increase the 
affected ECCS response times. The injection valves will still 
provide the function for which they were designed. Since the systems 
will continue to function as intended, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3) The proposed changes do not involve a significant reduction 
in a margin of safety because:
    While the calculated licensing basis PCT is larger than that 
previously calculated with the current valve stroke times, the new 
PCT remains far below the 2200 deg.F licensing acceptance limit 
associated with a LOCA. This limit has been previously evaluated as 
providing a sufficient margin of safety. All other LOCA licensing 
limits also continue to be met with the increased stroke times. For 
other accidents and transients, the increased stroke times have a 
negligible effect on the results, so the margin of safety is 
preserved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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

    Date of amendment request: April 15, 1992, as modified by letters 
dated December 8, 1992, and June 25, 1993.
    Description of amendment request: The proposed amendment would 
revise the provisions in the Technical Specifications to incorporate 
Generic Letter 90-06, ``Resolution of Generic Issue 70, `Power-Operated 
Relief Valve and Block Valve Reliability,' and Generic Issue 94, 
`Additional Low-Temperature Overpressure Protection for Light-Water 
Reactors,''' power-operated relief valve (PORV) requirements for power 
operation, and to modify the primary coolant system (PCS) overpressure 
protection specification venting requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification requiring Operability of 
the PORVs and their block valves does not alter plant operation or 
configuration in any way. It is current practice to maintain these 
valves in an Operable condition to meet the requirements of existing 
Specification 3.1.8, which is applicable when below 430 deg.F. The 
effect of the proposed changes is to extend the applicability of the 
Operability requirement for these valves. The addition of PORV 
Operability requirements when at Hot Standby and above would not 
involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    Replacing the requirements to vent the PCS through a 1.3 square 
inch vent with a vent capable of relieving 167 gpm at a pressure 
less than the Appendix G limits will not significantly increase the 
probability or consequences of an overpressurization event 
occurring. The 1.3 square inch vent area in Technical Specification 
3.1.8 was intended to be a means of protecting the Primary Coolant 
System (PCS) from exceeding the limit of the 10 CFR [Part] 50, 
Appendix G, curve following an overpressure transient. Analysis has 
shown that manual vent valves PC-514 and PC-515 will provide a 
relief capacity of 167 gpm at a PCS pressure of approximately 115 
psig, well below the minimum 331 psig limit (Appendix G curve limit 
for a 40 deg.F/hr heat-up). This relief capacity will protect the 
PCS against a pressure transient caused by a maximum charging/
letdown imbalance coincident with a 40 deg.F/hr PCS heat-up rate and 
a 60 deg.F/hr pressurizer heat-up rate.
    Two other pressure transients, a High Pressure Safety Injection 
(HPSI) pump start and a Primary Coolant Pump (PCP) start, are also 
precluded. With the PCS in a vented and depressurized state, the PCS 
would be below 212 deg.F Existing technical specifications require 
both HPSI pumps to be rendered inoperable below 260 deg.F and, with 
the system depressurized, normal operating procedures prohibit a PCP 
start due to insufficient pump net positive suction head (NPSH).
    Therefore, the 1.3 square inch requirement can be replaced with 
a requirement to have a vent capable of relieving 167 gpm at a PCS 
pressure less than or equal to the Appendix G limit with no 
significant increase in the probability or consequences of an 
overpressurization event occurring.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The addition of PORV Operability requirements when at Hot 
Standby and above will not alter plant operation or configuration. 
It will not alter any equipment or analyses. Therefore the addition 
of these PORV Operability requirements will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The new technical specifications requirements for PCS vent 
capacity will provide an equivalent or better, overpressure 
protection as compared to the existing requirement. No analysis has 
been found that shows that the existing 1.3 square inch vent area 
will protect the PCS from exceeding the Appendix G curve limit. 
However, analysis has been developed which shows that manual vent 
valves PC-514 and PC-515 will provide adequate relief capacity, 
maintaining PCS pressure within the 10 CFR [Part] 50, Appendix G, 
limits. Furthermore, two other pressure transients, a HPSI pump 
start and a PCP start, are also precluded by either existing 
technical specifications or normal plant operating procedures.
    Another related analysis has shown that relief valve RV-3164, 
the Low Pressure Safety Injection (LPSI) pumps, the LPSI pump seals, 
and the system piping of the shutdown cooling system have the 
capability of providing adequate overpressure protection to the 
shutdown cooling system.
    The addition of the manual vent valves do not introduce a vent 
path where a vent path had not previously existed. Therefore, the 
possibility of an accident of a new or different type, than 
previously evaluated in the FSAR, will not be created.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety will not be reduced by the proposed 
Technical Specifications changes. The extension of PORV Operability 
requirements has no effect on any margin of safety. The previous 
requirement assumed that a vent with an equivalent flow area as the 
original PORV would provide the same relief as the PORV itself and 
gave no consideration to how that flow area should factor in system 
losses or vent location. The new technical specification requirement 
offers a means to ensure the PCS will be protected against all 
achievable overpressure transients for the system configuration, 
with analyses to support it.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: A. Randolph Blough, Acting

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

    Date of amendment request: December 16, 1993
    Description of amendment request: The proposed amendment clarifies 
the requirements for maintaining secondary containment integrity when 
one or more Reactor Building Ventilation supply and exhaust valves are 
declared inoperable. The technical specification (TS) change adds a new 
Limiting Condition for Operation, Basis Statement, and Surveillance 
Requirements for these isolation valves. The change revises TS 
definition 1.14c, adds new Specifications 3.5.B.2, 3.5.B.3, 4.5.R, and 
a Basis statement to TS 3.5, edits T.S. 3.5.B.1.1. It also renumbers TS 
3.5.B.2 through 3.5.B.4, to 3.5.B.4 through 3.5.B.7. It also revises 
specification references within to reflect new specification numbers.
    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:
    GPU Nuclear has determined that operation of the Oyster Creek 
Nuclear Generating Station in accordance with the proposed Technical 
Specifications does not involve a significant hazard. The changes do 
not:
    1. Involve a significant increase in the probability or the 
consequence of an accident previously evaluated.
    The failure of any component in the [Reactor Building 
Ventilation System] RBVS was not considered as a credible initiating 
event for a design basis accident. However, the RBVS is designed to 
mitigate the consequences of a potential radiological release by the 
isolation of all supply and exhaust ducts to the environs. Since the 
failure of the RBVS was never considered as one of the initiators of 
an accident, this proposed change cannot increase the probability of 
occurrence of an accident. During the proposed Limiting Condition 
for Operation (LCO), the supply or exhaust duct will be isolated 
within 8 hours by one isolation valve secured in its post accident 
design position. Since the duct can perform its post accident design 
function (isolation), there is no increase in the consequences of an 
accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The design function of the RBVS automatic isolation valves is to 
isolate the ducts which penetrate the Reactor Building or Secondary 
Containment during a radiological release. During the LCO, the duct 
will be isolated within 8 hours by one isolation valve secured in 
the closed position. Since the duct will be isolated, this change 
will not create a possibility for an accident or malfunction of a 
different type than previously identified.
    3. Involve a significant reduction in a margin of safety.
    If a RBVS automatic isolation valve (supply) is declared 
inoperable, the proposed LCO would allow continued plant operation 
with that supply duct isolated. Since the RBVS can still perform its 
design function (redundant ductwork) under normal plant and design 
accident conditions, there is no reduction in the margin of safety. 
For an inoperable isolation valve in the exhaust duct, the exhaust 
duct will be isolated within 8 hours by one isolation valve secured 
in the closed position. Further, the RBVS and the [Standby Gas 
Treatment System] SGTS will be aligned for an accident condition 
with no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment requests: November 17, 1993
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications to allow a portion of the Waste Gas 
Holdup System (WGHS) Explosive Monitoring System to be inoperable for 
160 days on a one-time basis. This is to allow replacement of the Waste 
Gas Oxygen Analyzer.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    We have evaluated the proposed T/S changes and have determined 
that the changes should involve no significant hazards 
consideration. Operation of the Cook Nuclear Plant in accordance 
with the proposed amendment will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The purpose of the hydrogen and oxygen monitors is to 
measure the concentrations of these gases in the WGHS to ensure that 
the gas mixture is non-flammable. We can accomplish this purpose and 
ensure safe operation of the WGHS by operating the system in the 
proposed manner. During the equipment replacement interval, we will 
be assuming that the hydrogen concentration is above the flammable 
limit (4%). The limiting factor then is the oxygen concentration at 
which hydrogen and oxygen become flammable. The existing hydrogen 
monitor will be in continuous operation to verify the hydrogen 
level. The information provided by the oxygen monitor being replaced 
is not essential to the safe operation of the WGHS since it is 
redundant to information provided by the remaining oxygen monitor. 
The only difference between the two (2) oxygen monitors is that the 
one being replaced provides an automatic isolation of the waste gas 
decay tank when the oxygen concentration reaches 3%. The isolation 
of the waste gas decay tanks will be performed manually during the 
replacement. In the event the remaining oxygen monitor becomes 
inoperable, we will follow the currently approved T/Ss. Since 
operation of the WGHS in the manner we have proposed will ensure 
that the purpose of the oxygen and hydrogen monitors is fulfilled 
and safe operation of the WGHS is maintained, the proposed change 
will not involve a significant increase in the probability or 
consequences of a previously analyzed incident.
    The proposed change to the Automatic Gas Analyzer (QC-31) tag 
number to QC-1400 will not reduce in any way requirements or 
commitments in the existing T/Ss. The proposed change will eliminate 
confusion of spare parts of the new analyzer panel installed in 
1990.
    (2) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any previously evaluated. During 
the replacement of the monitors the waste gas holdup system will 
continue to operate normally. The proposed method of operation will 
ensure that the oxygen and hydrogen gas mixture is non-flammable. 
For this reason, operating the explosive gas monitoring system in 
the proposed manner will not place the plant in a new or unanalyzed 
condition. Therefore, we believe that this change will not introduce 
a new or different kind of accident than previously analyzed.
    The proposed editorial change will not create the possibility of 
a new or different kind of accident from any previously evaluated, 
because these changes will not place the plant in a new or 
unanalyzed condition.
    (3) Involve a significant reduction in a margin of safety.
    The proposed amendment does not involve a significant reduction 
in the margin of safety. The remaining oxygen monitor will be 
available to maintain the oxygen concentration below the limit 
required for hydrogen flammability in oxygen. In addition, the 
oxygen grab samples will provide redundant information and will 
serve as a check of the monitor's readings. If the remaining monitor 
becomes inoperable, we will follow the actions of our current T/Ss. 
During the equipment replacement period, we will be assuming that 
the hydrogen concentration is above the flammable limit (4%). This 
will then make the oxygen level the controlling parameter in a 
possible flammable combination of oxygen and hydrogen. The existing 
hydrogen monitor will be in continuous operation to verify the 
hydrogen level. These proposed interim measures will not 
significantly affect our ability to maintain the hydrogen and oxygen 
concentration within the limits to prevent flammability. Therefore, 
we believe that operation of the system in this manner does not 
involve a significant reduction in a margin of safety.
    The proposed editorial change will not involve a significant 
reduction in margin of safety, because all accident analyses and 
nuclear design bases remain unchanged.
    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: A. Randolph Blough, Acting

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 requests: December 20, 1993
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications to change Train A and B Emergency 
Loads from 8 hour to composite 4 hour, delete a load on the Train B 
batteries load list, and revise the operational loads on the Train N 
batteries.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Per 10 CFR 50.92, a proposed amendment does not involve a 
significant hazards consideration if the change does not:
    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) create the possibility of a new or different kind of 
accident from any previously analyzed.
    (3) involve a significant reduction in a margin of safety.
    Criterion 1
    The change is proposed to comply with the LOCA/LOOP [loss-of-
coolant accident/loss-of-offsite power] and SBO [Station Blackout] 
requirement for Cook Nuclear Plant for battery testing. The 
composite test as addressed above meets these requirements for four 
hour test profiles. This change is consistent with the UFSAR three 
hour LOCA/LOOP and NUMARC 87-00 Station Blackout Rule four hour. 
Based on these considerations, the proposed change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Criterion 2
    The change only addresses the battery profile test which meets 
both LOCA/LOOP and SBO for Cook Nuclear Plant. No specific physical 
or operational changes to the plant will occur due to this change. 
Thus, the change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Criterion 3
    The proposed change revises the battery profile test table from 
an eight hour to a four composite test which complies with both 
LOCA/LOOP and SBO as defined for Cook Nuclear Plant. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: A. Randolph Blough, Acting

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 requests: December 22, 1993
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications section addressing steam generator 
stop valves making it more consistent with the revised Standard 
Technical Specifications and clarifying certain surveillance 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Per 10 CFR 50.92, a proposed change does not involve a 
significant hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    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.
    Criterion 1
    The limiting conditions for operation involving the steam 
generator stop valves are not altered by this proposed change. The 
surveillance requirements are lessened for Unit 2 in that valve 
stroke timing does not have to be performed on valves that are 
closed. This is consistent with the wording of the Unit 1 T/S, and 
reflects the fact that, when closed, the valves are already in the 
position required by the assumptions in the safety analysis and 
therefore stroke timing is not necessary. The remaining changes are 
consistent with NUREG 1431, and as such, have already been found 
acceptable by the NRC. Therefore, it is concluded that the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Criterion 2
    No changes to the limiting conditions for operation of the steam 
generator stop valves are proposed as part of this amendment 
request. The proposed changes do not involve any physical changes to 
the plant. The changes will allow operation in Modes 2 and 3 with 
more than one steam generator stop valve inoperable. However, 
inoperable valves must be closed and their closure periodically 
reverified. When closed, the valves are already in the position 
required by the assumptions in the safety analysis. Thus, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Criterion 3
    The limiting conditions for operation involving the steam 
generator stop valves are not altered by this proposed change. The 
surveillance requirements are lessened for Unit 2 in that valve 
stroke timing does not have to be performed on valves that are 
closed. This is consistent with the wording of the Unit 1 T/S, and 
reflects the fact that, when closed, the valves are already in the 
position required by the assumptions in the safety analysis and 
therefore stroke timing is not necessary. The remaining changes are 
consistent with NUREG 1431, and, as such, have already been found 
acceptable by the NRC. Therefore, it is concluded that the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: A. Randolph Blough, Acting

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

    Date of amendment request: December 22, 1993
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.4.e (Emergency Ventilation 
System). TS 3.4.4.e currently permits fuel handling operations to 
continue during refueling for up to 7 days with one circuit of the 
emergency ventilation system inoperable provided all active components 
of the other emergency ventilation system circuit are operable. The 
proposed revision would permit fuel handling operations to continue 
during refueling beyond 7 days with one circuit of the emergency 
ventilation system inoperable provided the remaining emergency 
ventilation system circuit is operable and in operation. The licensee 
stated that the proposed revision is consistent with recently issued 
Amendment No. 47 to the Nine Mile Point Unit 2 TSs and with the NRC's 
Improved Standard Technical Specifications, NUREG-1433.
    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 [NMP1], in accordance 
with the proposed amendment, will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Emergency Ventilation System responds to a release of 
radioactivity to the secondary containment by maintaining a negative 
pressure in secondary containment and by providing a filtered 
elevated release. The proposed change to LCO [Limiting Condition for 
Operation] 3.4.4.e would allow continuation of refueling beyond 
seven days with one emergency ventilation circuit inoperable 
provided the operable emergency ventilation circuit is in operation. 
A plant specific PRA [Probabilistic Risk Assessment] was performed 
to evaluate the probability of a fuel bundle drop event resulting in 
a need to start the Emergency Ventilation System with a concurrent 
failure of the Emergency Ventilation System that would result in an 
unfiltered ground level release under the current Technical 
Specifications and the proposed change. The results of this 
assessment indicate that the probability is not significantly 
increased. In addition, the order of magnitude of the probability of 
such a release, under the current or proposed Technical 
Specifications, is very small, i.e., 10-6. This amendment 
requires no physical changes to NMP1. Therefore, the proposed 
changes to the Technical Specifications do not significantly 
increase the probability of an accident previously evaluated.
    Section XV.C.3 of the UFSAR [Updated Final Safety Analysis 
Report] evaluates a fuel bundle drop accident. The radiological 
consequences of this accident are within the guidelines of 10 CFR 
Part 100. The UFSAR radiological evaluation takes credit for the 
operation of an emergency ventilation circuit in mitigating the 
consequences of this accident. During refueling with one emergency 
ventilation circuit inoperable for more than seven days, the 
proposed Technical Specification change would require that an 
operable emergency ventilation circuit be placed in operation. With 
an operable emergency ventilation circuit operating prior to a fuel 
bundle drop accident, the radiological consequences of this accident 
remains bounded by the current UFSAR evaluation. Therefore, from a 
radiological perspective, the proposed Technical Specification 
change is bounded by the current radiological evaluation in the 
UFSAR. Therefore, the Technical Specification change does not 
significantly increase the consequences of a previously evaluated 
accident.
    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.
    This amendment does not involve any accident precursors or 
initiators. During an accident involving the release of 
radioactivity to the secondary containment atmosphere an operable 
emergency ventilation circuit would already be running and 
performing its safety function. The operating status of a running 
emergency ventilation circuit, which was manually started, would be 
unaffected by the receipt of an automatic start signal due to the 
detection of high radiation in secondary containment. Accordingly, 
the proposed Technical Specification change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The current Technical Specifications, LCO 3.4.4, provide a 
margin of safety by requiring both emergency ventilation circuits to 
be operable during a refueling condition. With one emergency 
ventilation circuit inoperable, the current Technical Specifications 
allow continuation of refueling for up to seven days, at which time 
refueling must be stopped. This Technical Specification requirement 
ensures that an emergency ventilation circuit will be available to 
provide a filtered release to the environment during an accident 
which could result in the release of radioactivity to the secondary 
containment atmosphere.
    The proposed change to LCO 3.4.4.e would allow continuation of 
refueling beyond seven days with one emergency ventilation circuit 
inoperable provided the operable emergency ventilation circuit is in 
operation. By placing the remaining operable emergency ventilation 
circuit in operation, active single failures associated with its 
startup have been eliminated. These eliminated failures include 
automatic initiation instrumentation, relaying logic, breaker 
operation, fan startup and valve operation. With an operable 
emergency ventilation circuit in operation, its safety function is 
being performed. In addition, the status of the operating emergency 
ventilation circuit is indicated in the control room. Therefore, the 
running, operable emergency ventilation circuit provides a level of 
safety comparable to two non-running, operable emergency ventilation 
circuits.
    Based upon the above analysis, the margin of safety is not 
significantly reduced by allowing refueling to continue beyond seven 
days with one emergency ventilation circuit inoperable since the 
operable emergency ventilation circuit is in operation.
    These changes are consistent with Amendment No. 47 for Nine Mile 
Point Unit 2 and with the Improved Standard Technical 
Specifications, NUREG-1433.
    Accordingly, as determined by the analysis above, this proposed 
amendment involves no significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 27, 1993
    Description of amendment request: The proposed amendment would 
relocate Technical Specification (TS) Tables 3.2.7, ``Reactor Coolant 
Isolation Valves,'' and 3.3.4, ``Primary Containment Isolation 
Valves,'' from TS 3.2.7/4.2.7 and 3.3.4/4.3.4, respectively, to a plant 
procedure overning lists removed from TSs per Generic Letter (GL) 91-
08, ``Removal of Component Lists from Technical Specifications.'' The 
plant procedure would be subject to the requirements specified in the 
Administrative Controls section of the Nine Mile Point Nuclear Station 
Unit No. 1 (NMP-1) TS. The proposed amendment would also make 
conforming changes to the TS Bases. These lists of valves will continue 
to be included in the NMP-1 Updated Final Safety Analysis Report 
(FSAR). The licensee stated that the proposed changes would be 
consistent with NRC staff guidance issued in GL 91-08.
    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.
    1.The proposed amendment does not involve a physical change to 
any system, structure, or component that affects the probability or 
consequences of any accident or malfunction of equipment important 
to safety.
    Relocation of the component lists to plant procedures and the 
Updated FSAR is in accordance with Generic Letter 91-08. This change 
does not alter the existing operability or surveillance requirements 
for the components to which they apply. The proposed changes are 
under the change control provisions in Section 6.0, ``Administrative 
Controls,'' of the Technical Specifications. The changes associated 
with the Bases for Specification[s] 3.2.7/4.2.7 and 3.3.4/4.3.4 are 
consistent with the issuance of prior license amendments. Since the 
proposed amendment does not affect the operation or testing of any 
plant systems or components, it will have no impact on the 
probability or consequences of accidents or malfunctions previously 
evaluated.
    2.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 changes to Technical Specification 3.2.7/4.2.7, 
3.3.4/4.3.4 and Bases do not introduce any new modes of plant 
operation or new accident precursors, involve any physical 
alterations to plant configurations, or make changes to system 
setpoints which could initiate a new or different kind of accident. 
The proposed changes relocate Reactor Coolant Primary Containment 
Isolation Valve Tables 3.2.7 and 3.3.4 into a procedure governing 
controlled lists removed from TS per Generic Letter 91-08 under the 
change control provisions in Section 6.0, ``Administrative 
Controls,'' of the Technical Specifications. The testing associated 
with these valves remains unchanged, therefore, it will not affect 
system or component operability. In addition, the removal of generic 
reference to the 60 second closure time is consistent with 
previously issued license amendments and has no impact on either the 
Limiting Condition for Operation or Surveillance Requirement. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The operation of Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The Technical Specification Limiting Conditions for Operation 
and Surveillance Requirements for the valves listed in Tables 3.2.7 
and 3.3.4 are not being altered. The valve lists will be 
incorporated into a procedure governing controlled lists removed 
from TS per Generic Letter 91-08. This is controlled by Section 6.0, 
``Administrative Procedures.''
    In addition, removal of generic reference to the 60 second 
closure time is consistent with previously issued license amendments 
and has no impact on either the Limiting Conditions for Operation or 
Surveillance Requirements. Therefore, the proposed changes will not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: October 4, 1993
    Description of amendment request: The proposed amendment would 
delete the emergency diesel-generator engine speed specification from 
Surveillance Requirement (SR) 4.8.1.1.2a.5 and would replace the diesel 
engine speed requirement with an electrical frequency requirement in SR 
4.8.1.1.2g. Currently, SR 4.8.1.1.2a.5 specifies both a minimum engine 
speed and a nominal electrical frequency and acceptable deviation from 
the nominal value. SR 4.8.1.1.2g currently specifies only a minimum 
engine speed. The specified minimum engine speed is not consistent with 
the acceptable frequency deviation below the nominal value.
    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.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)) because the proposed changes merely delete the 
specification of emergency diesel-generator minimum engine speed 
while retaining or substituting the specification of electrical 
frequency to be attained by the emergency diesel-generator. The 
diesel engine and generator are directly coupled and must rotate at 
the same speed, therefore, speed and frequency are directly related 
and specification of acceptance values for both parameters is 
redundant. Furthermore, electrical frequency, not engine speed, is 
the parameter of interest with regard the ability of the system to 
power emergency loads. The proposed changes do not affect the 
adequacy of the surveillance test or the reliability of the system 
to power emergency loads, and do not involve any physical changes to 
facility structures, systems, or components. Therefore, since the 
reliability of the emergency diesel-generators will not be reduced, 
the probability or consequences of any accident previously evaluated 
is not increased.
    B. The changes do not create the possibility of a new or 
different kind
    of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because no physical changes to facility structures, 
systems, or components are involved and they do not affect the 
manner by which the facility is operated.
    C. The changes do not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes 
do not affect the manner by which the facility is operated or 
involve changes to equipment or features which affect the 
operational characteristics of the facility. 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: Exeter Public Library, 47 
Front Street, Exeter, New Hampshire 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston Massachusetts 02110-2624.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: October 28, 1993
    Description of amendment request: The proposed amendment would 
implement 13 of the 47 line item Technical Specification (TS) 
improvements recommended by Generic Letter 93-05. Most of the proposed 
changes would revise the allowable time intervals for performing 
certain Surveillance Requirements (SR) on various plant components 
during power operation or would delete the requirement entirely or 
under certain conditions. One proposed change would modify testing 
requirements identified in an ACTION statement. The specific changes 
are as follows:
    1. SR 4.1.3.1.2 - The allowable interval between tests to 
demonstrate the operability of any partially or fully withdrawn control 
rod would be increased to 92 days from 31 days.
    2. SR 4.6.4.1 - The allowable interval between tests to demonstrate 
the operability of the hydrogen monitors by performing an Analog 
Channel Operational Test (ACOT) would be increased to 92 days from 31 
days, and by performing a Channel Calibration to every refueling outage 
from 92 days on a staggered basis.
    3. SR 4.3.2.1, Table 4.3-2, Functional Unit 3.c.4 and SR 4.3.3.1, 
Table 4.3-3, Functional Units 1 through 6 - The allowable interval 
between tests to demonstrate the operability of the radiation monitors 
by performing an ACOT and Digital Channel Operational Test (DCOT) would 
be increased to quarterly from monthly.
    4. SR 4.4.6.2.2- The time the plant may be in Cold Shutdown before 
Pressure Isolation Valve testing is required prior to entering Mode 2 
would be increased to 7 days from 72 hours.
    5. SR 4.4.11.1 - The allowable interval between tests to 
demonstrate the operability of the Reactor Coolant System vent block 
valves would be increased to cold shutdown from every 92 days.
    6. SR 4.4.3.2 - The allowable interval between tests to verify 
pressurizer heater capacity would be increased to each refueling outage 
from 92 days.
    7. SR 4.5.1.1.1 - The requirement to verify the boron 
concentration of the accumulator contents after a volume increase of 
1-percent or more would be removed under certain conditions.
    8. SR 4.5.1.1.2 - The requirement to perform an ACOT and Channel 
Calibration on accumulator water level and pressure instrumentation 
would be deleted.
    9. SR 4.5.2 - The requirement to visually inspect the containment 
sump upon completion of each containment entry would be modified to 
avoid unnecessary containment sump inspections when multiple 
containment entries are made on the same day.
    10. SR 4.6.2.1 - The allowable interval between tests to verify 
that each containment spray nozzle is unobstructed would be increased 
to every 10 years from 5 years.
    11. SR 4.6.4.2 - The allowable interval between tests to 
demonstrate operability of each hydrogen recombiner system would be 
increased to each refueling interval from 6 months.
    12. SR 4.7.1.2.1 - The allowable interval between tests of the 
auxiliary and startup feedwater pumps would be increased to 92 days on 
a staggered test basis from 31 days.
    13. TS 3.8.1.1 - The ACTION statements would be changed so that 
when the Limiting Conditions for Operation are not met due to:
    a. Inoperability of one or two offsite power circuits the starting 
of a diesel-generator would no longer be required;
    b. Inoperability of one diesel-generator, starting of the remaining 
diesel-generator would be required within 8 hours only under certain 
conditions instead of within 24 hours under all conditions;
    c. Inoperability of a diesel-generator and an offsite power 
circuit, starting of the remaining diesel-generator would be required 
within 8 hours only under certain conditions instead of within 24 hours 
under all conditions.
    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.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)) because the proposed changes either merely modify the 
allowable intervals between certain surveillance tests, delete the SR, 
or alter an ACTION statement with regard to required testing. The 
safety functions of the related structures, systems, or components are 
not changed in any manner nor are the reliabilities of any structure, 
system, or component reduced by the revised surveillance or testing 
requirements. The changes do not affect the manner by which the 
facility is operated and do not change any facility design feature, 
structure, system, or component. Since there is no change to the 
facility or operating procedures, and the safety functions and 
reliabilities of structures, systems, or components are not affected, 
there is no affect upon the probability or consequences of any accident 
previously analyzed.
    B. The changes do not create the possibility of a new or different 
kindof accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because they do not change the facility or affect the 
manner by which the facility is operated. The proposed changes merely 
change certain surveillance or testing requirements.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because they do not affect the manner by 
which the facility is operated or change equipment or features which 
affect the operational characteristics of the facility.
    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: Exeter Public Library, 47 
Front Street, Exeter, New Hampshire 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston Massachusetts 02110-2624.
    NRC Project Director: John F. Stolz

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: December 17, 1993
    Description of amendment request: The proposed amendment changes 
the action statements for the limiting conditions associated with the 
electrical power sources (Technical Specification 3.8.1.1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    In accordance with 10 CFR 50.92, NNECO has reviewed the attached 
proposed changes and has concluded that they do not involve a 
significant hazards consideration. The basis for this conclusion is 
that the three criteria of 10CFR50.92(c) are not compromised. The 
proposed changes do not involve a significant hazards consideration 
because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to rewrite the action statements for 
Technical Specification 3.8.1.1 will decrease the wear on the EDGs 
[emergency diesel generators] by reducing the number of required 
starts. These changes will also allow adequate time for the 
completion of manufacturer recommended EDG engine prelube and warmup 
procedures. They ensure that the operability of the AC sources is 
demonstrated with reasonable assurance. Also, the reliability of the 
EDGs will be enhanced by reducing the potential for numerous 
unloaded EDG starts during an LCO [limiting condition for operation] 
period. The proposed changes could reduce the number of required 
unloaded EDG starts from nine to one. Therefore, these proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to rewrite the action statements for 
Technical Specification 3.8.1.1 do not alter the method of operating 
the plant, nor do they introduce any new failure modes. The proposed 
changes affect EDG testing frequency only, they have no impact on 
any accident analysis. The proposed changes provide assurance that 
the EDGs will be able to power their respective safety systems if 
required. Also, they do not involve any physical alterations to 
plant equipment or procedures which would introduce any new or 
unique operational modes or accident precursors.
    3.Involve a significant reduction in a margin of safety.
    The proposed changes to rewrite the action statements for 
Technical Specification 3.8.1.1 do not affect the capability of the 
EDGs to perform their function. The intent of the changes is to 
increase the overall EDG reliability, by reducing the wear resulting 
from excessive and unwarranted testing. 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: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: June 12, 1992, as supplemented September 
17, 1992, March 17, 1993, August 17, 1993, August 18, 1993, and 
December 29, 1993.
    Description of amendment request: The purpose of the proposed 
changes is to revise the Technical Specifications (TS) to permit 
operation of the James A. FitzPatrick Nuclear Power Plant at an uprated 
power of 2536 MWt. The licensee's engineering analyses and evaluations 
confirm that the plant can be operated at an uprated power. The 
increase in the rated power from 2436 MWt to 2536 MWt corresponds to a 
4.8 percent increase in rated steam flow. The increase in rated power 
remains below the plant design power level of 2550 MWt which was the 
basis for the original plant safety evaluation.
    The changes affect the operating parameters of the reactor, 
operational restrictions, setpoints for safety systems, analytical 
results, and test requirements. There are also administrative changes. 
The changes in each of these categories are summarized as follows:
    Reactor Parameters: The effect on reactor parameters is limited. 
Higher power is achieved by control rod pattern adjustments to increase 
reactor thermal power in a more uniform (flattened) powerdistribution 
to increase steam flow without increasing core recirculation flow. This 
requires an increased reactor dome pressure for adequate turbine inlet 
pressure.
    Operational Limits: The increased thermal power requires a change 
to the limitation on operation in the high power low flow portion of 
the power/flow map to limit thermal hydraulic instabilities and power 
oscillations.
    Setpoints: The increased reactor pressure has a direct impact on 
the high pressure scram setpoint and the safety relief valve setpoint. 
Additionally, the bypass for the turbine stop valve closure and control 
valve fast closure scram will be changed in proportion to the increase 
in thermal power.
    Analysis Results: Analyses of uprated power transients and 
accidents requires changes to various TS and their Bases. Operational 
parameters and assumptions used in analyses were revised to reflect 
their use as initial conditions. Revised radiological analyses changed 
dose results. The results of the accident analyses requires revisions 
to properly reflect plant capabilities.
    Testing: A number of changes to testing requirements result from 
power uprate. The increase to reactor pressure has a direct effect on 
hydrostatic leakage testing pressure. The test pressure for High 
Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling 
(RCIC) pumps will be revised to reflect safety relief valve setpoints 
assumed in analyses.
    Administrative: Administrative changes (i.e., adding references, 
revising references and correcting associated errors) will also be 
made.
    No changes to the Radiological Effluent Technical Specifications 
were identified.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant at a thermal power of 2536 
MWt will not involve a significant hazards consideration as defined 
in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The James A. FitzPatrick nuclear power plant was reviewed for 
operation at a rated power of 2550 MWt at the time of its operating 
license, [***]. This review was based on the original design of the 
plant. Since that time, a number of safety issues of a generic and 
plant specific nature as well as plant modifications have changed 
the originally reviewed design.
    Generic criteria, methodologies and evaluation scope required to 
uprate BWRs up to 5% were prepared by General Electric and submitted 
to the NRC in LTR-1 [NEDC-31897P-A, ``Generic Guidelines for General 
Electric Boiling Water Reactor Power Uprate]. This was 
supplemented by the submittal of generic evaluations in LTR-2 [NEDC-
31984P, ``Generic Evaluations of General Electric Boiling Water 
Reactor Power Uprate'' and Supplement 1] to determine: which NRC and 
industry generic communications were applicable to power uprate and 
how they should be treated; analytical evaluations that could be 
generically approved; bounding evaluations of components and 
equipment, and; the effect of power uprate on safety margin. These 
generic evaluations are supplemented by plant specific evaluations. 
The Power Uprate Safety Analysis Report (PUSAR) describes the 
dependence placed on References 1 and 2, the additional analyses 
that were performed, the results of these additional analyses and 
overall conclusions on the safety impacts of power uprate.
    The plant systems and components will be within design limits at 
power uprate conditions with minor modifications. At uprated power, 
the power plant will not be operated in a manner that is different 
from current operations except for limited changes to operating 
parameters such as primary system pressure, steam flow and feedwater 
temperature. Setpoints are revised as necessary to reflect new 
operational conditions and analyses. The ECCS-LOCA [emergency core 
cooling system-loss-of-coolant accident] analysis using current 
practices demonstrates compliance with design and regulatory 
acceptance criteria at uprated power.
    The radiological consequences of accidents have been evaluated 
using more current methodologies with consistent assumptions and 
continue to meet acceptance criteria. Compliance with NRC dose 
criteria using current methodologies is discussed in Section 9.2 of 
the PUSAR. The effect of power uprate on dose analyses now discussed 
in the FSAR [Final Safety Analysis Report] were qualitatively 
assessed recognizing that power uprate increases doses in direct 
proportion to the 4.1% increase in thermal power. An increase of 
4.1% to the calculated doses currently identified in FSAR Chapter 14 
indicates that a reevaluation using the original methodology would 
have demonstrated compliance with current NRC dose criteria. A 
review of Table 14.4-2 indicates that, with the 4.1% increase, 
offsite doses would be substantially less than NRC allowable values. 
A review of Table 14.8-1 indicates that, with the 4.1% increase, 
control room doses would be substantially less than NRC allowables 
except for the main steam line break (MSLB). However, the MSLB dose 
would drop well below allowables once the proposed change on 
allowable coolant activity (reduces the limit by more than a factor 
of ten) is accounted for.
    [Thus, based on the above analysis and supplemental analyses 
performed by the NRC staff, the increase in power level discussed 
herein and associated Technical Specification changes do not 
significantly increase 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.
    Operation at uprated power involves no changes to the manner in 
which the plant is operated. There are changes to operational 
parameters and setpoints but analyses of these identified no new 
failure modes or accident scenarios. The effects of transients and 
accidents fall within design capabilities. Systems and components 
are capable of operating and performing their safety functions at 
uprated power. No mechanisms for creating a new or different 
accident were identified.
    3. involve a significant reduction in a margin of safety.
    The power uprate will not result in significant increases to 
primary system temperature and pressure due to postulated operating 
transients or accidents. These and other margins of safety have been 
discussed in the PUSAR, where it is demonstrated that there will be 
no reductions in the margin of safety because the plant will still 
meet its design and regulatory acceptance criteria. For example, the 
core will continue to be operated with the same margin to the safety 
limit minimum critical power ratio. Fuel thermal limits will 
continue to meet NRC acceptance criteria. Plant systems and 
equipment are designed for uprated power conditions and have been 
evaluated for their capability to perform at uprated conditions. 
They will continue to perform within design limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 22, 1993
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications proposes to remove the 
reference to American Society for Testing and Materials (ASTM) Standard 
D 270-65 from Surveillance Requirement 4.12.A.1.i. ASTM D 270-65, which 
specifies procedures to draw a representative fuel oil sample, has been 
superseded and is no longer in effect. The FitzPatrick Surveillance 
Procedure currently requires sampling in accordance with ASTM D 270-65 
but notes that it has been superseded by ASTM D 4057-88. The proposed 
change makes Surveillance Requirement 4.12.A.1.i consistent with fuel 
oil Surveillance Requirement 4.9.C.1, for the emergency diesel 
generators (EDGs), by adopting the current industry fuel oil sampling 
standard. There are no changes to the acceptance criteria for fuel oil 
quality which are based on ASTM D 975-81.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Plant accident analyses are not affected by the Technical 
Specification change. The change removes reference to ASTM D 270-65 
as the method for obtaining samples of diesel fuel oil from 
Surveillance Requirement 4.12.A.1.i. The American Society for 
Testing and Materials has superseded ASTM D 270-65 with ASTM D 4057. 
The proposed change provides improved flexibility to adopt standards 
as they are issued without requiring a license amendment and makes 
Surveillance Requirement 4.12.A.1.i consistent with both 
Surveillance Requirement 4.9.C.1 for the EDG and the STS [Standard 
Technical Specification]. The nature of this change will not cause 
any increase in the probability or consequences of previously 
evaluated accidents.
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed change involves no hardware modifications to any 
plant structures, systems or components. The change removes 
reference to ASTM D 270-65 as the method for obtaining samples of 
diesel fuel oil from Surveillance Requirement 4.12.A.1.i. The 
American Society for Testing and Materials has superseded ASTM D 
270-65 with ASTM D 4057. The nature of this change is such that no 
new or different kind of accident can be created.
    3. involve a significant reduction in the margin of safety.
    The proposed change will not cause a reduction in the margin of 
safety. The results of the plant accident analyses continue to bound 
operation under the proposed changes so there is no reduction in the 
margin of safety. The change removes reference to ASTM D 270-65 to 
allow the use of ASTM D 4057-88 as the method for obtaining samples 
of diesel fire pump fuel oil. The change will make Surveillance 
Requirement 4.12.A.1.i consistent with existing Surveillance 
Requirement 4.9.C.1 for the EDG and allow current standards to be 
used for fuel oil sampling. Revisions of this nature will not cause 
a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied.
    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 28, 1993
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications (TSs) revises the scope 
of startup and power escalation reports to reflect the guidance 
provided by the Standard Technical Specifications (STS). Current TS 
requirements state, ``The report shall address each of the tests 
identified in the FSAR [Final Safety Analysis Report] ....''
    The list of tests provided in the FSAR includes many tests which 
were required for the initial plant startup but are not performed for 
subsequent startups. Currently, the licensee satisfies TS requirements 
in the startup reports by identifying tests which are not performed, 
their purpose, and the reason for not performing them.
    The inclusion of such a section in the startup reports is 
unnecessary. This amendment would revise the TSs to permit subsequent 
startup reports to address only those tests that are actually 
performed.
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve no hardware modifications, no 
changes to the operation of any system or component, no changes to 
structures, and alters future startup reports by addressing only 
tests that are performed. A change in reporting format will not 
eliminate the performance of startup tests that are necessary or 
required. These changes can not alter the probability or 
consequences of existing accident analyses as documented in the FSAR 
or the NRC staff SER [Safety Evaluation Report].
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The changes do not alter the testing procedures or 
methodologies. The changes provide a basis, in accordance with NRC 
guidance, for allowing future startup test reports to mention and 
discuss only those tests that are actually performed. These changes 
can not result in a new or different type of accident than those 
previously evaluated.
    3. involve a significant reduction in the margin of safety.
    There are no changes to tests that are performed in support of 
plant startup [***]. Removing a requirement for reporting on tests 
that are not performed will not cause a reduction in any margin of 
safety.
    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 NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 29, 1993
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications would eliminate an 
inconsistency between the Reactor Coolant System (RSC) leakage 
detection and operability requirements in Limiting Conditions for 
Operation (LCO) 3.6.D.1 and 3.6.D.4.
    LCO 3.6.D.1 currently requires that RCS leakage be below specified 
limits when there is irradiated fuel in the reactor vessel and reactor 
coolant temperature is greater than 212 deg.F. LCO 3.6.D.4 requires the 
leakage monitoring systems to be operable during power operation (i.e., 
when the mode switch is in the Startup/Hot Standby position or the Run 
position with the reactor critical above 1 percent rated power, as 
defined per Specification 1.0.O). These two LCOs are not consistent. 
The proposed revision of LCO 3.6.D.4 will take the more conservative 
approach of requiring the leakage monitoring systems to be operable 
when the leakage limits of LCO 3.6.D.1 are in effect.
    The proposed changes also make editorial corrections which are 
considered administrative in nature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment involves no hardware changes, no changes 
to the operation of any systems or components and no changes to 
structures. It alters an LCO to require plant leakage detection 
systems to be operable during the same plant conditions that RCS 
leakage limits apply. The revised LCO requires leakage detection 
systems in operational modes consistent with other portions of the 
Technical Specifications. Additional changes include editorial 
corrections such as correct specification numbering, proper system 
identification, and clarification of a surveillance requirement 
cross reference. Since the change to the LCO for leakage detection 
will require operability under a greater range of plant conditions 
to be consistent with detection requirements, there is no change to 
previously determined accident probabilities or consequences. The 
editorial changes have no adverse safety implications.
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed amendment involves no hardware changes, no changes 
to the operation of any systems or components and no changes to 
structures. It alters the Technical Specifications only to the 
extent of making two LCOs consistent by requiring the leakage 
detection system to be operable when leakage limits apply and making 
editorial changes. Editorial changes and increasing the plant 
conditions for leakage system operability can not create the 
possibility of a new or different kind of accident from those 
previously evaluated since the editorial changes have no safety 
significance and the operability changes are being made for 
consistency with the modes when leakage detection is required to 
function.
    3. involve a significant reduction in the margin of safety.
    The proposed amendment revisions involve no hardware changes, no 
changes to the operation of any systems and no changes to 
structures. The revised LCO criteria for RCS leakage detection 
system operability has increased the plant conditions when 
operability is required to match the plant conditions when leakage 
limits apply. Editorial changes and expanded operability 
requirements in the LCOs will not result in any change to existing 
safety margins.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 29, 1993
    Description of amendment request: This proposed amendment to the 
James A. FitzPatrick Technical Specifications, Appendix B, revises 
Surveillance Requirement 3.1.a and Table 3.10-2. The existing 
Surveillance Requirement in Specification 3.1.a, which references Table 
3.10-2, requires the performance of instrument checks, tests, and 
calibrations to assure the operability of specific gaseous effluent 
radiation monitors. It does not require any surveillances for the 
associated data recorders. However, Limiting Condition for Operation 
(LCO) 3.1.a specifically states that ''...pathways shall be monitored 
and recorded ....'' which requires the recorders to be operable for the 
gaseous effluent monitoring system to be considered operable. The 
proposed changes to the Radiological Effluent Technical Specifications 
(RETS) add the radiation monitor recorders to the Surveillance 
Requirement and Table 3.10-2 for the gaseous effluent monitoring 
system. This will provide the surveillance requirements for the data 
recorders.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment involves no hardware changes, no changes 
to the operation of any systems or components and no changes to 
structures. It alters procedures by including the recorders in the 
Surveillance Requirements for determining the operability of the 
radiological monitoring/recording instrumentation. The inclusion of 
the data recorders in the Surveillance Requirements for checking, 
testing, and calibration does not reduce gaseous effluent monitoring 
capability while providing greater assurance that data is recorded. 
Adding the existing system surveillance to the Technical 
Specifications will not have any affect on previously evaluated 
accident probabilities or consequences.
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed amendment involves no hardware changes, no changes 
to the operation of any systems or components and no changes to 
structures. It alters procedures by including the data recorders in 
the Surveillance Requirements for the radiation monitors. These 
changes do not affect the manner in which the gaseous effluent 
radiation monitoring system is operated or tested. The resulting 
changes do not pose a safety issue concern different from those 
analyzed previously in the FSAR [Final Safety Analysis Report] or 
the NRC staff SER [Safety Evaluation Report].
    3. involve a significant reduction in the margin of safety.
    The proposed amendment involves no hardware changes, no changes 
to the operation of any systems and no changes to structures. It 
alters procedures by including the data recorders in the 
Surveillance Requirements for the radiation monitors. The revised 
surveillance requirement increases the scope of surveillance for the 
gaseous effluent radiation monitoring system, for consistency with 
the LCO. It does not cause any reduction in any safety margins.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: January 11, 1994
    Description of amendment request: This proposed amendment to the 
James A. FitzPatrick Technical Specifications (TS) would temporarily 
waive the 2-year maximum surveillance interval for the Type C test 
(local leak rate test) of the shutdown cooling isolation valves (10MOV-
17 and 10MOV-18). The waiver would permit deferring this test until the 
next refueling outage, currently scheduled for the end of November 
1994.
    TS 4.7.A.2.e(5) and 10 CFR Part 50 Appendix J require the 
containment isolation valves to be leak-rate tested during each 
refueling outage, but at intervals of not greater than 2 years. As a 
result of the extended 1991-1993 refueling outage, and the length of 
the current operating cycle, tests of all containment isolation valves 
are due prior to the start of the next refueling outage. Two mid-cycle 
outages were scheduled, in part, to accommodate the testing of these 
valves, except for the shutdown cooling isolation valves. The shutdown 
cooling isolation valves cannot be removed from service during a non-
refueling outage to accommodate a leak-rate test since its associated 
system is needed to remove reactor decay heat. The reliability of the 
isolation valve design, and the very low probability that the shutdown 
cooling system penetration would result in a pathway for leakage to the 
reactor building, as discussed in the application, is provided as 
justification for this one-time schedular extension.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is limited to a one-time schedular extension 
in the shutdown cooling isolation valve Type C test. The change does 
not introduce any new modes of plant operation, make any physical 
changes, or alter any operational setpoints. Therefore, the change 
does not degrade the performance of any safety system assumed to 
function in the accident analysis. The extension will not result in 
a significant increase in valve leakage considering that: (1) the 
valves are infrequently stroked, and then only when the reactor 
vessel is at low pressure, (2) monitoring the normal operating 
status of the RHR [residual heat removal] system assures the absence 
of gross valve leakage, and (3) the valves were replaced with valves 
of an improved design which has been confirmed by past Type C tests 
to exhibit satisfactory leak rate performance. For these reasons, 
the change does not involve a significant increase in the 
probability of an accident.
    The change does not involve a significant increase in the 
consequences of an accident evaluated since any leakage through the 
shutdown cooling penetration will not significantly increase for 
reasons discussed in the previous paragraph, and such leakage is 
negligible compared to the main steam line break accident analyzed 
in the FSAR [final safety analysis report].
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed change does not introduce new accident initiators 
or failure mechanisms since the change does not alter the physical 
characteristics of any plant system or component. The change is 
limited to a one-time schedule extension for the shutdown cooling 
isolation valve Type C tests.
    3. involve a significant reduction in the margin of safety.
    There is a very low probability of a significant increase in 
valve leakage considering the demonstrated reliability of the 
current valve design, the infrequent use of the valves, and the 
monitoring of the normal operating status of the RHR system. 
Moreover, any potential incremental benefit of performing the tests 
within the two year requirement would not be sufficient to offset 
the increased occupational radiation exposure associated with 
testing, and the risk to plant safety associated with the removal 
from service of the primary method of decay heat removal. 
Consequently, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: October 1, 1993 TS 93-09
    Description of amendment request: The proposed changes would revise 
the setpoints and time delays for the auxiliary feedwater (AFW) and 6.9 
kv shutdown board loss of voltage and degraded voltage instrumentation. 
The proposed changes would affect Technical Specification (TS) Tables 
3.3-3, 3.3-4, 3.3-5, 4.3-2, and the Bases for Specification 3/4.7.1.2. 
Table 3.3-3 would be revised: (1) to reflect the use of a two-out-of-
three voltage sensor logic for loss of power detection (and one-out-of-
two logic scheme for the timing relays) by adding requirements for the 
respective AFW, 6.9 kv shutdown board, and emergency diesel generator 
(EDG) voltage sensors and load shed timers; (2) by changing the 
description of the functional unit from ``Station Blackout'' to ``Loss 
of Power Start;'' (3) by adding a footnote to indicate that the new 
requirements apply only to the shutdown boards on the same unit; (4) by 
increasing the modes for which the EDG sensors and timers must be 
operable to include Modes 5 and 6 when the associated EDG must be 
operable; (5) by changing the associated Action statements to reflect 
the design changes and consistency; (6) to reflect consistent 
terminology in the Table and action statements; (7) by changing the 
minimum number of channels that would be required to be operable; (8) 
by changing the footnote to reflect the conditions when the loss-of-
power instrumentation is required to be operable in Modes 5 and 6; and 
(9) by adding an exclusion to Specification 3.0.4. Similar changes were 
proposed for Tables 3.3-4, 3.3-5 and 4.3-2. The proposed change to 
Bases 3/4.7.1.2 would clarify AFW operability on loss of 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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revision supports the implementation of design 
logic and setpoint changes to the loss-of-power relaying. This 
relaying is designed to ensure adequate voltage is available to 
safety-related loads in order to enhance their operability and 
support accident mitigation functions and to provide for auxiliary 
feedwater (AFW) pump starts. The design changes alter relay logic 
and delete unnecessary relaying, but do not change the diesel 
generator (D/G) start and load-shedding actuations that result from 
loss-of-power conditions. Therefore, no new actuations or functions 
have been created; and because the existing and proposed functions 
provide for accident mitigation considerations that are not the 
source of an accident, the probability of an accident is not 
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feeder undervoltage relays actually reduces the potential for 
inadvertent shutdown board blackouts as a result of short-duration 
voltage transients or instrument failures.
    The setpoints and time delays for loss-of-power functions have 
been modified based on the guidelines developed by the Electrical 
Distribution System Clearinghouse as evaluated and determined 
through detailed analysis by TVA. This design is documented in TVA 
Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is 
available for NRC review at the SQN site. The assigned values are 
conservative settings that will ensure adequate voltage is supplied 
to safety-related loads for accident mitigation and safety functions 
under normal, degraded, and loss-of-offsite-power voltage conditions 
with appropriate time delays to prevent damage to electrical loads 
and minimize premature or unnecessary actuations. The identification 
of loss-of-voltage conditions is enhanced by the design changes to 
ensure the timely sequencing of loads onto the D/G and the 
initiation of AFW pump starts for accident mitigation. Because there 
are no reductions in safety functions resulting from the design 
logic, setpoint, and time-delay changes to the loss-of-power 
instrumentation and offsite dose levels for postulated accidents 
will not be increased, the consequences of an accident are not 
increased.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification incorporated in the proposed 
change do not affect plant functions. These changes reflect the 
requirements that SQN has been maintaining and serve to clarify the 
requirements to provide consistency of application and easier 
understanding. The AFW footnote addition and bases revision only 
clarify operability conditions that are consistent with the plant 
design for the AFW pump and loss-of-power instrumentation. Because 
there are no changes to plant functions or operations, these 
revisions have no impact on accident probabilities or consequences.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As described above, the loss-of-power instrumentation ensures 
adequate voltage to safety-related loads by initiating D/G starts 
and load shedding and provides for AFW pump starting, but is not 
considered to be the source of an accident. Although the design 
logic, setpoint, and time-delay actuation criteria have changed, the 
output functions to various plant systems that actuate for load 
shedding and D/G starts remain the same. Therefore, actuation 
criteria have been affected, but not safety functions, and the TVA 
evaluation has confirmed that the new design enhances the ability to 
maintain adequate voltage to support safety functions. Since safety 
functions have not changed and the new loss-of-power instrumentation 
design continues to support operability of safety-related equipment, 
no new or different accident is created.
    The applicable mode addition, TS 3.0.4 exclusion deletion, and 
response time measurement clarification, as well as the AFW 
operability clarifications, do not affect plant functions and will 
not create a new accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed loss-of-power TS changes support design logic, 
setpoint, and time-delay requirements that have been verified by TVA 
analysis to provide acceptable voltage levels for safety-related 
components. In determining the acceptability of these voltage 
levels, the minimum voltage for operation as well as detrimental 
component heating resulting from sustained degraded-voltage 
conditions were considered. This design ensures that safety-related 
loads will be available and operable for normal and accident plant 
conditions. The applicable mode addition, TS 3.0.4 exclusion 
deletion, response time measurement clarification, and AFW 
operability clarifications provide enhancements to TS requirements 
and do not affect plant functions. Therefore, no safety functions 
are reduced by these changes and there is no reduction in the margin 
of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: November 18, 1993 (TS 93-17)
    Description of amendment request: The proposed change would delete 
the requirement of License Condition 2.H of the Operating License. This 
license condition requires reporting the violation of certain license 
conditions to the NRC Regional Administrator within 24 hours by phone 
and facsimile, and a followup report within 14 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is a deletion of an administrative reporting 
requirement that does not in any way affect a previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Since there are no changes in the way the plant is operated, the 
potential for an unanalyzed accident is not created. The proposed 
change is administrative in nature and does not affect any accident 
initiators for SQN. No new failure modes are introduced.
    3.Involve a significant reduction in a margin of safety.
    Plant safety margins are established through limiting conditions 
of operation, limiting safety system settings, and safety limits 
specified in the TSs. As a result of the proposed amendment, there 
will be no changes to either the physical design of the plant or to 
any of these settings and limits. The proposed changes are 
administrative and do not affect the safe operation of SQN. 
Therefore, there will be no changes to any of the margins of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: December 7, 1993 (TS 93-16)
    Description of amendment request: The proposed change would 
incorporate the following changes to Section 6.0, ``Administrative 
Controls,'' of the Technical Specifications (TS): (1) delete Section 
6.1.2 that indicates that the Corporate Manager of Radiological Control 
has the responsibility for the radiological environmental program, dose 
calculations, and projections; (2) added a requirement in Section 
6.5.1.6 for review of the Offsite Dose Calculations Manual (ODCM) by 
the Plant Operations Review Committee (PORC); (3) added implementation 
of the ODCM to Section 6.8.1 as an activity requiring written 
procedures and deleted the requirement that they be maintained by the 
Radiological Control Group; (4) delete the requirement in Section 
6.8.4.a for Radiological Control to implement and control the ODCM; (5) 
changed the review authority for changes to the ODCM from the 
Radiological Assessment Review Committee (RARC) to the PORC and added a 
reference to Specification 6.5.1A in Section 6.14.1.2; (6) approval 
authority for deviations from the overtime guidelines in Section 
6.2.2.g would be changed to show that Plant Manager designee also has 
the authority; (7) PORC member titles would be changed in accordance 
with the current organizational structure; (8) the requirements for the 
RARC would be deleted from Sections 6.5.2.7.i, 6.5.3, 6.10.2.k and the 
index; (9) move the condenser inleakage monitoring requirement from 
Section 6.8.5.c.(vii) to 6.8.5.c.(iii) and delete the requirement to 
repair, plug, or isolate leaks; (10) change the title ``Shift 
Supervisor'' (SS) to ``Shift Operations Supervisor'' (SOS) in various 
locations in Section 6 and in Operating License Items 2.C.(23).A and 
2.C.(16).a for Units 1 and 2 respectively; (11) change the title of the 
Senior Vice President, Nuclear Group, in Section 6.2.1 to Senior Vice 
President, Nuclear Power; (12) change the title of the Operational 
Quality Assurance Program in Section 6.5.2.8.d to Nuclear Quality 
Assurance Program; and (13) move the requirement for implementation of 
the Quality Assurance Program for environmental monitoring from Section 
6.8.4.b to Section 6.8.1.h. Other administrative changes related to 
these changes were also submitted that affect Section 6.0, the 
operating license, index, and definitions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). The operation of Sequoyah Nuclear Plant (SQN) in 
accordance with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes only affect the administrative controls 
found in Section 6.0 of the SQN TSs and the operating license. No 
plant equipment or operating practices are affected by these 
changes. The revised administrative controls will continue to 
adequately implement administrative activities to support plant 
nuclear safety. Since there are no physical changes to the plant, 
there is no increase in the probability of an accident because these 
administrative controls are not the source of previously evaluated 
accidents. Similarly, with no change to plant equipment or operating 
requirements, the plant response to accident conditions and 
therefore the consequences of an accident remain unchanged. These 
proposed changes will not increase the consequences of an accident 
and offsite dose rates will not be impacted.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The administrative controls affected by the proposed changes are 
not considered to be the source of any accident and these changes 
will not alter any plant features or processes. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident and the administrative controls will 
continue to implement the actions necessary to support plant 
activities and nuclear safety.
    3. Involve a significant reduction in a margin of safety.
    Plant features and setpoints remain unchanged by the proposed 
changes to the administrative controls. The margins of safety 
established by the SQN design are not affected by these changes. The 
proposed administrative controls will continue to maintain the 
actions and programs that ensure appropriate plant design, 
operation, and procedures to support the required margin of safety. 
Therefore, the proposed changes will not reduce the margin of 
safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: September 29, 1993
    Description of amendment request: The proposed amendments would 
change the test intervals from monthly to quarterly, consistent with 
the Inservice Test Program, for several pumps and related systems, 
including safety injection, residual heat removal and containment spray 
pumps.
    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 amendments will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The addition of specific general considerations related to 
equipment surveillance requirements and their relationship to 
equipment operability are administrative in nature. They do not 
change the interpretation or intent of the Technical Specifications. 
These conditions are consistent with the Westinghouse Standard 
Technical Specifications (STS).
    The addition of the specific requirement for the Inservice Test 
Program to Section 15.4.2 of the Technical Specifications only 
reiterates the requirements of 10 CFR 50.55a(g). This requirement 
does not implement any new requirements on the operation or testing 
of equipment.
    The decrease in the number of equipment operational transients 
due to the increase in the surveillance interval for the Safety 
Injection System (SI), Residual Heat Removal (RHR) System, and 
Containment Spray (CS) System pumps and valves will result in an 
increase in system availability. Reduced testing is also expected to 
have a positive affect on overall equipment reliability since 
frequent testing results in increased wear and potential for 
equipment failure. Other actions including a monthly verification of 
system lineups for the SI, CS and RHR systems provides increased 
assurance of system operability between surveillance tests. The 
potential for equipment problems to go undetected for a longer 
period of time is small as indicated by equipment surveillance 
history.
    Therefore, these changes will not effect the probability or 
consequences of previously analyzed accidents.
    2. The proposed amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    These changes only affect the equipment testing frequency. 
Equipment design, operation and the methods of testing will not be 
changed. Therefore, the proposed changes cannot create the 
possibility of a new or different kind of accident than any accident 
previously evaluated.
    3. The proposed amendments will not involve a significant 
reduction in the margin of safety.
    The proposed changes which implement the requirements of 10 CFR 
50.55a(g) and clarify the general considerations related to 
equipment surveillances are administrative in nature. They do not 
change the intent of any existing license or other requirement.
    The increase in equipment surveillance intervals will result in 
an improvement in equipment and system availability and reliability. 
Surveillance of equipment will be performed as required by the 
regulations and Section XI of the ASME Boiler and Pressure Vessel 
Code. These proposed changes will reduce the potential for equipment 
failures due to unnecessary testing.
    Adequate assurance is provided by testing in accordance with the 
ASME Code requirements and periodic verification of system lineups 
to ensure that the affected systems remain operable and capable of 
performing their design function. Therefore, a reduction in a margin 
of safety will not occur.
    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    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: December 1, 1993
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) by incorporating technical and administrative changes to TS 3.10, 
Control Rod and Power Distribution Limits. The proposed changes 
eliminate specifications for fuel designs no longer used at Kewaunee, 
specify required actions to be taken upon exceeding control bank 
insertion limits, and revise the limits for Departure from Nucleate 
Boiling (DNB) related parameters to assure operation within the 
assumptions of the Updated Safety Analysis Report (USAR) analyses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (a) TS 3.10.b.1, 3.10.b.4 and Table TS 3.10-2
    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.
    The proposed changes eliminate the specifications for fuel that 
is no longer used at the Kewaunee Nuclear Power Plant. Eliminating 
these specifications does not affect the probability of any accident 
previously evaluated.
    The specifications for the current fuel vendor are being 
retained and ensure the consequences of previously evaluated 
Departure from Nucleate Boiling (DNB) related accidents are 
enveloped by the Updated Safety Analysis Report (USAR) analyses. 
Therefore, these changes will not increase the 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.
    The proposed change deletes specifications for fuel that is no 
longer used at the Kewaunee Nuclear Power Plant. The limits for the 
current fuel vendor are retained and are not affected by this 
proposed change. This does not alter the input or assumptions of the 
safety analysis, and therefore it will not involve a reduction in 
the margin of safety.
    (b) TS 3.10.d
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change specifies the required actions to be taken 
when the control bank insertion limits are exceeded. The current TS 
requires compliance with the control bank insertion limits, but 
gives no corrective action for when these limits are exceeded. The 
new specification requires operators to initiate boration to restore 
shutdown margin within one hour of exceeding the control bank 
insertion limits and to restore the control banks to within the 
limits within 2 hours. If either of these requirements cannot be 
achieved, within 1 hour the operators must initiate actions to 
achieve Hot Standby within the next 6 hours and Hot Shutdown within 
the following 6 hours. Adding these requirements clarifies and 
enhances the Technical Specifications and will have no impact on the 
probability of an accident previously evaluated.
    The proposed addition is conservative and ensures that proper 
and adequate measures are taken when the control bank exceeds the 
control bank insertion limits. Therefore, this addition will not 
increase the 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.
    The proposed change is conservative and clarifies the necessary 
actions to be taken when control bank insertion limits are exceeded. 
This proposed change is an enhancement to the specification and does 
not reduce the margin of safety.
    (c) TS 3.10.k
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change decreases the maximum RCS inlet temperature 
limit for steady state 100% operation from 536.5 degrees Fahrenheit 
to 535.5 degrees Fahrenheit. The value of 539.5 degrees Fahrenheit 
is the assumed RCS inlet temperature for the DNB related accidents 
analyzed in the USAR. These accidents are the Uncontrolled Rod 
Cluster Control Assembly Withdrawal at Power Accident, the 
Uncontrolled Rod Cluster Control Assembly Withdrawal from a 
Subcritical Condition Accident, the Rod Cluster Control Assembly 
Misalignment, the Start-Up of an Inactive Reactor Coolant Loop 
Accident, the Excessive Heat Removal due to Feedwater System 
Malfunction Accident, the Excessive Load Increase Incident, the Loss 
of Reactor Coolant Flow Accident, the Loss of External Electrical 
Load, the Steam Line Break, and the Rod Cluster Control Assembly 
Ejection. A four degree assumed instrument error reduces the maximum 
allowed RCS inlet temperature to 535.5 degrees Fahrenheit. 
Decreasing this value to ensure consistency with the USAR analysis 
assumptions will have no impact on the probability of an accident 
previously evaluated.
    The proposed change is conservative to ensure that the 
consequences of a previously evaluated DNB-related accident are 
enveloped by the USAR analysis. Therefore, this change will not 
increase the 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.
    The proposed change is conservative and is consistent with the 
assumptions in the USAR. This proposed change is an enhancement to 
the specification and does not reduce the margin of safety.
    (d) TS 3.10.l
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change increased the minimum RCS pressure limit for 
steady-state 100% power operation from the currently specified 2200 
psig to 2205 psig. The value of 2205 psig (30 psig below the nominal 
design value of 2235 psig) was the assumed initial condition for the 
DNB related accidents analyzed in the USAR. These accidents are the 
Uncontrolled Rod Cluster Control Assembly Withdrawal at Power 
Accident, the Uncontrolled Rod Cluster Control Assembly Withdrawal 
from a Subcritical Condition Accident, the Rod Cluster Control 
Assembly Misalignment, the Start-Up of an Inactive Reactor Coolant 
Loop Accident, the Excessive Heat Removal due to Feedwater System 
Malfunction Accident, the Excessive Load Increase Incident, the Loss 
of Reactor Coolant Flow Accident, the Loss of External Electrical 
Load, the Steam Line Break, and the Rod Cluster Control Assembly 
Ejection. Increasing this value to ensure consistency with the USAR 
analysis assumptions will have no impact on the probability of an 
accident previously evaluated.
    The proposed change is conservative to ensure that the 
consequences of a previously evaluated DNB-related accident is 
enveloped by the USAR analysis. Therefore, this change will not 
increase the 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.
    The proposed change is conservative and is consistent with the 
assumptions in the USAR. This proposed change is an enhancement to 
the specification and does not reduce the margin of safety.
    (e) TS 3.10.m
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1)Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    TS 3.10.m.1 provides the limits and required actions to be taken 
when the RCS flow rate per loop is less than the USAR analysis 
assumptions. Decreasing the flow limit to be consistent with the 
USAR assumptions will have no impact on the probability of an 
accident previously evaluated. Compliance with the flow limit 
assumed in the USAR analyses ensures the consequences of previously 
evaluated DNB-related accidents are enveloped by the USAR analyses. 
Therefore, this change will not increase the consequences of an 
accident previously evaluated in the USAR.
    The proposed revision places an additional restriction on RCS 
flow when less than the design flow rate of 89,000 gallons per 
minute per loop. The intent of specifying action in accordance with 
3.10.n is to ensure that reactor power is reduced to a point at 
which the DNB ratio margin is restored. Compliance with this 
specification will not increase the probability of an accident 
previously evaluated, nor increase the consequences of an accident.
    The intent of TS 3.10.m.2 is to clarify the conditions under 
which the reactor coolant flow rate is verified. The conditions, 
methodology, and uncertainties associated with this verification are 
not changed by this specification. Clarifying the TS by inclusion of 
the conditions for verifying the flow rate will not increase the 
probability or consequences of any accident previously evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accidents previously evaluated.
    A new or different kind of accident from those previously 
evaluated will not be created by this TS change. The proposed 
amendment does not alter the plant configuration, operating 
setpoints or overall plant performance.
    3) Involve a significant reduction in the margin of safety.
    A specification on steady-state RCS flow rate is necessary to 
ensure DNB ratio criteria will be met during the DNBR limiting 
events analyzed in the USAR. Reducing the reactor coolant flow limit 
to the value assumed in the USAR analyses does not result in a 
reduction in the margin of safety.
    The additional restrictions being imposed if the specified limit 
is not met provide additional assurance the DNBR margin will be 
restored. These additional restrictions do not exist in the current 
TS. The imposition of these restrictions results in an enhancement 
to the margin of safety.
    Clarifying the TS by inclusion of conditions under which the 
flow verification is to be performed will not reduce the margin of 
safety. Existing approved constraints, methodology, and 
uncertainties are not being changed by this clarification.
    (f) TS 3.10.n
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The intent of this new TS is to outline the actions required 
when the limits of TS 3.10.k (RCS temperature), TS 3.10.l (RCS 
pressure) and TS 3.10.m.1 (RCS flow) are exceeded. Collectively, 
these three specifications place limits on the DNB-related 
parameters to assure each is maintained within the normal steady-
state envelope assumed in the USAR safety analysis. This 
specification is an enhancement to our existing specification to add 
clear guidance which does not presently exist. Providing this 
information for the plant staff and operators will not increase the 
probability of an accident previously evaluated.
    The addition of this action statement will ensure that the 
consequences of an analyzed accident are not increased. The proposed 
specification allows 2 hours to evaluate and restore parameters to 
within limits. If this time frame is not satisfied, then within the 
next 6 hours, power is reduced in order to restore a margin of 
safety. Following analysis, thermal power may be raised not to 
exceed a level analyzed to maintain a minimum DNBR of 1.30.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment does not alter the plant configuration, 
operating setpoints or overall plant performance.
    3) Involve a significant reduction in the margin of safety.
    Addition of the specification is an enhancement to the current 
specification which does not alter input to the safety analysis. 
Therefore, it will not involve a reduction in the margin of safety.
    (g) Administrative changes to Section TS 3.10 including Figure 
TS 3.10-2.
    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.
    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: December 7, 1993
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
5.3.a.1 to provide flexibility in the repair of fuel assemblies 
containing damaged and leaking fuel rods by reconstituting the 
assemblies, provided that an NRC-approved methodology is used. This 
proposed change is consistent with guidance provided in Supplement 1 to 
Generic Letter (GL) 90-02, ``Alternative Requirements for Fuel 
Assemblies in the Design Features Section of Technical 
Specifications,'' dated July 21, 1992. In addition, administrative 
changes to KNPP TS Section 5 have been proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    The proposed changes were revised in accordance with the 
provision 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.
    This proposed change to the requirements for ``Fuel Assemblies'' 
in the ``Design Features'' section of the KNPP TS will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. This proposed change will not result 
in modifications to fuel assemblies that would have a significant 
effect on safety because of the requirement to implement these 
changes using an NRC-approved methodology. This requirement will 
confirm conformance to existing design limits and confirm that 
safety analyses criteria are met before operation during the next 
fuel cycle. This license amendment request is consistent with 
guidance provided by the NRC and will result in flexibility for 
improved fuel performance.
    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The creation of new or different kind of accident from any 
previously evaluated accident is not considered a possibility 
because the changes are administrative in nature and do not 
represent an actual modification to the plant or change its safety 
analyses.
    3) involve a significant reduction in the margin of safety.
    The margin of safety is maintained by adherence to other fuel 
related TS limits and the USAR design bases. The changes do not 
directly affect any safety system or the safety limits, and thus 
does not affect the plant margin of safety.
    Accordingly, these proposed changes 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: 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

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.

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

    Date of application for amendments: October 19, 1993
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) add a footnote to TS 4.6.1.2.b that 
allows a one-time exemption from the accelerated containment integrated 
leak rate test (CILRT) requirements to return the CILRT frequency for 
both units to a normal Type A test interval.
    Date of issuance: January 11, 1994
    Effective date: January 11, 1994
    Amendment Nos.: 167 and 198
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59745) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application of amendments: May 3, 1993, as supplemented 
August 11, 1993.
    Brief description of amendments: The amendments revise the limiting 
conditions for operation and surveillance requirements related to the 
Low Pressure Service Water System.
    Date of issuance: January 13, 1994
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.:  203, 203, and 200
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52983) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 13, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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

    Date of application for amendment: October 27, 1993
    Brief description of amendment: The amendment relocated the 
requirement of Technical Specification 4.5.2.g.1 to verify the correct 
position of each electrical and/or mechanical position stop for the 
Emergency Core Cooling System throttle valves within 4 hours of each 
valve stroking operation or maintenance on the valve, to procedures 
that control the maintenance and operation of these valves.
    Date of issuance: January 14, 1994
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment No.:  155
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64606) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 14, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801

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

    Date of application for amendment: August 11, 1993
    Brief description of amendment: The amendment changed the Grand 
Gulf Nuclear Station Technical Specifications to support compliance 
with the new requirements of Title 10 Code of Federal Regulations Part 
20 and Part 50.36a. The request to change the wording of TS 1.46 which 
relates to the definition of an UNRESTRICTED AREA remains under 
consideration and will be the subject of a future licensing action.
    Date of issuance: January 10, 1994
    Effective date: January 10, 1994
    Amendment No: 111
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46233) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 10, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, Post Office Box 1406, S. Commerce at Washington, Natchez, 
Mississippi 39120.

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

    Date of amendment request: August 27, 1993.
    Description of amendment request: The amendment changes the 
footnote on page 1 of License NPF-86 by deleting Vermont Electric 
Generation and Transmission Cooperative, Inc., (Vermont), as one of the 
entities for which North Atlantic Energy Service Corporation (North 
Atlantic) is authorized to act. The change reflects the purchase of 
Vermont's share of the Seabrook Station, Unit 1 by North Atlantic 
Energy Corporation (NAEC) pursuant to a prior settlement of a claim by 
Vermont against Public Service Company of New Hampshire (PSNH). NAEC 
acquired PSNH's interest in the Seabrook Station, Unit 1 in accordance 
with the Plan for Reorganization for PSNH.
    Date of issuance: January 7, 1994
    Effective date: To be implemented by May 30, 1994.
    Amendment No.:  28
    Facility Operating License No. NPF-86. Amendment revised the 
License.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52990). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 7, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, New Hampshire 03833.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: October 7, 1992 as supplemented July 12, 
1993
    Brief description of amendment: The amendment changed the setpoint 
limit for the degraded-voltage protection system referred to as the 
offsite-power low signal.
    Date of issuance: January 14, 1994
    Effective date: January 14, 1994
    Amendment No.:  159
    Facility Operating License No. DPR-40. Amendment revised the 
TechnicalSpecifications.
    Date of initial notice in Federal Register: November 25, 1992 (57 
FR 55584) The additional information contained in the supplemental 
letter dated July 12, 1993, was clarifying in nature and, thus, within 
the scope of the initial notice and did not affect the staff's proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated January 14, 1994. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: September 15, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) by implementing Generic Letters (GLs) 86-10 and 
88-12. This removed the fire protection TSs and placed these parts in 
the Updated Safety Analysis Report (USAR).
    Date of issuance: January 14, 1994
    Effective date: January 14, 1994

    Amendment No.:  160
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59753) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 14, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendment: August 3, 1993
    Brief description of amendment: This amendment removes shutdown 
system control valves and primary containment isolation valves from TS 
Tables 3.3.7.4-1, ``Remote Shutdown Instrumentation and Controls,'' and 
3.6.3-1, ``Primary Containment Isolation Valves,'' as a result of 
eliminating the steam condensing mode of the Residual Heat Removal 
system.
    Date of issuance: January 12, 1994
    Effective date: January 12, 1994
    Amendment No. 65
    Facility Operating License No. NPF-39. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50969) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 12, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: May 21, 1993, as supplemented 
October 7, 1993, and December 3, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to incorporate the following changes:
    (1) The safety injection system test frequency (specified in TS 
Section 4.5.A.1.a) was changed to accommodate operation on a 24-month 
cycle.
    (2) The loss of normal AC in conjunction with a safety injection 
signal test frequency (specified in TS Section 4.6.A.3) was changed to 
accommodate operation on a 24-month cycle. This TS section was also 
reformatted to improve clarity.
    (3) The auxiliary feedwater system undervoltage automatic start 
test frequency (specified in TS Table 4.1-1) was changed to accommodate 
operation on a 24-month cycle.
    (4) The auxiliary feedwater system main feedwater pump trip 
automatic start test frequency (specified in TS Table 4.1-1) was 
changed to accommodate operation on a 24-month cycle.
    In addition, quarterly testing and 24-month calibration 
requirements were added to TS Table 4.1-1 for the main steam line flow 
instrumentation. These surveillances were added to ensure operability 
of the main steam line flow circuits and to be consistent with the TSs 
surveillance requirements for other engineered safety features 
instruments.
    These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle.''
    Date of issuance: January 11, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.:  142
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41510)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 11, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: October 29, 1993
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) Sections 3.10 (Control Rods and Power Distribution 
Limits) and 4.2 (Inservice Inspections) to correct administrative 
errors that resulted from the issuance of TS Amendment Nos. 57 and 103. 
The amendment corrects the errors and further clarifies the TS.
    Date of issuance: January 12, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 143
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64615) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 12, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: October 30, 1991
    Brief description of amendment: The amendment revised Technical 
Specification 3.1.3.2, ``Control Rod Maximum Scram Insertion Times,'' 
to clarify the conditions under which the plant must be shut down in 
the event that individual control rod scram insertion times exceed the 
allowable values.
    Date of issuance: January 19, 1994
    Effective date: January 19, 1994
    Amendment No. 54
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 11, 1991 (56 
FR 64651) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 19, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By March 4, 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).

Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear 
Power Station, Unit 2, Rock Island County, Illinois

    Date of application for amendments: October 29, 1993, as 
supplemented by letters dated December 22, 1993 and January 14, 1994.
    Brief description of amendments: The license amendment dispositions 
Unreviewed Safety Questions (USQ) related to proposed plant 
modifications associated with reactor vessel water level 
instrumentation. These modifications have been initiated to mitigate 
the circumstances outlined in NRC Bulletin 93-03, ``Resolutions of 
Issues Related to Reactor Vessel Water Level Instrumentation in BWRs.''
    Date of issuance: January 19, 1994
    Effective date: January 19, 1994
    Amendment No.: 139
    Facility Operating License No. DPR-30. 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 January 19, 
1994.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    NRC Project Director: James E. Dyer
    Dated at Rockville, Maryland, this 26th day of January 1994.
    For the Nuclear Regulatory Commission.
John N. Hannon,
Acting Director, Division of Reactor Projects - III/IV/V, Office of 
Nuclear Reactor Regulation
[Doc. 94-2174 Filed 2-1-94; 8:45 am]
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