[Federal Register Volume 61, Number 177 (Wednesday, September 11, 1996)]
[Notices]
[Pages 47973-47987]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20911]


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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

[[Page 47974]]

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 August 19, 1996, through August 29, 1996. 
The last biweekly notice was published on August 28, 1996 (61 FR 
44353).

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By October 11, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any 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,

[[Page 47975]]

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 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: August 1, 1996
    Description of amendments request: The amendment will allow use of 
blind flanges during MODES 1-4 in the Calvert Cliffs Units 1 and 2 
Containment Purge Systems. These flanges will establish integrity in 
Mode 5, prior to entering Mode 4, and maintain it in Modes 1-4, 
functions presently served by the valve.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The purpose of the Containment Purge System is to provide 
ventilation for the containment while in a shutdown condition. 
Valves, which are disabled in the shut position in Modes 1-4, may be 
opened in Modes 5 and 6 to allow air flow, are provided in the 
supply and exhaust piping, and are automatically shut on a 
Containment Radiation Signal to prevent release of radioactive 
material in the event of a fuel handling incident. Manual operation 
is also provided. In Modes 1-4, the valves are kept shut to provide 
containment integrity to withstand a presumed increase in 
containment pressure in the event of a loss-of-coolant accident. The 
proposed change will allow blind flanges to serve in place of the 
purge valves in Modes 1-4 by blocking off the purge penetration on 
both the supply and exhaust sides. The blind flanges will provide 
the same level of containment integrity previously provided by the 
purge valves. The revised Technical Specifications will continue to 
verify containment building leakage is maintained within the 
allowable limits by requiring the performance of a 10 CFR Part 50, 
Appendix J, Type B, leakage test on the blind flanges. The outside 
valve in each containment purge penetration will be removed and the 
inside valves will be left in place. The remaining inside valves 
will no longer by required to provide containment integrity in Modes 
1-4. Only one of each pair of valves was credited for containment 
closure (Modes 5 and 6); therefore, removing the outside valves and 
the associated automatic closure signals is not a modification of 
the required capability to close the penetration. The inside valves 
will maintain their current safety function to close containment (if 
needed) by closing either on a Containment Radiation Signal (Mode 6) 
or manually (Modes 5 and 6). The Technical Specification 
surveillances associated with the purge valves will be changed to 
reflect the proposed modification to the plant. Since the blind 
flanges will limit radiological releases in Modes 1-4, and the purge 
valves will limit radiological releases in Modes 5 and 6, the 
proposed change will not increase the consequences of an accident 
previously evaluated.
    The Containment Purge System is not an accident initiator but 
acts to limit the consequences of accidents. The system will provide 
containment isolation in Modes 1-4 as before, and the inside valves 
will still be available to close in Modes 5 and 6. Therefore, the 
proposed change does not increase the probability of an accident 
previously evaluated.
    As stated above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    This requested change does not involve a significant alteration 
of the operation of the plant, and no new accident initiation 
mechanism is created by the modification. Four purge valves per unit 
currently provide containment closure in Modes 5 and 6. The outside 
valve in the supply and the exhaust lines will be removed to allow 
for installation of a blind flange in each line. The remaining 
supply and exhaust valves inside containment will continue to 
provide containment closure. The function currently performed by the 
four purge valves in Modes 1, 2, 3 and 4 will be performed by the 
blind flanges. Other, similar, blind flanges have been in service in 
the plant for a number of years, and have proven reliable. The 
Technical Specification surveillances associated with the testing of 
the purge valves and flanges will be changed to reflect the proposed 
modification to the plant. Therefore, this change does not create 
the possibility of a new or different type of accident from any 
accident previously evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    The valves in the Containment Purge System currently provide 
containment integrity during Modes 1, 2, 3 and 4, and containment 
closure during Modes 5 and 6. The function currently performed by 
the purge valves in Modes 1, 2, 3 and 4 will be performed by the 
blind flanges. Because of their design and mounting method, the 
blind flanges will perform the containment integrity function as 
well as, or better than, the purge valves. In Modes 1-4, the double 
o-rings in the blind flanges will provide single-failure protection 
similar to the other existing Type B penetrations. The established 
allowable containment building leakage rate will be maintained by 
the implementation of a requirement to perform 10 CFR Part 50, 
Appendix J, Type B, leakage rate on the installed blind flanges. The 
outside valve in each purge containment penetration will be removed. 
Single failure is not assumed in the fuel handling accident 
analysis, therefore, removing the outside valves and their 
Containment Radiation Signal channels is not a modification of the 
required capability to close the penetration. The remaining inside 
valves will continue to provide automatic and manual containment 
closure in Mode 6 to mitigate the effects of a fuel handling 
accident. The Technical Specification surveillances associated with 
purge valve testing will be changed to reflect the proposed 
modification to the plant. Therefore, this change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request

[[Page 47976]]

involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

    Date of application for amendment: August 7, 1996
    Brief description of amendment: The amendment proposes revising the 
Technical Specifications (TSs) to allow the use of 10 CFR Part 50, 
Appendix J, Option B, Performance-Based Containment Leakage Rate 
Testing. This performance-based Option B may be used as an alternative 
to the requirements in Appendix J, ``Primary Reactor Containment 
Leakage Testing for Water-Cooled Power Reactors,'' of 10 CFR Part 50. 
To implement Option B to Appendix J, the amendment proposes modifying 
TSs to eliminate reference to the prescriptive Appendix J requirements 
and instead reference NRC Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program.'' The amendment also proposes an 
editorial correction to the mathematical formula minimum testing 
frequency in the basis for TS 4.1.
    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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    For Indian Point Unit No. 2, the integrated leak rate testing 
[ILRT] as-found measured leakage rate acceptance criteria is changed 
from 0.75 La to 1.0 La. This change is consistent with the revised 
10 CFR 50 Appendix J, NEI 94-01, ``Industry Guidelines for 
Implementing Performance-Based Option of 10 CFR Part 50, Appendix 
J.'' In addition, an as-found leakage rate acceptance criteria of 
1.0 LA for Type A tests is consistent with the design basis and 
accident analysis assumptions. The as-left acceptance criteria 
remains unchanged at 0.75 La in accordance with the NEI guidance. 
Therefore, prior to entering an operating mode where containment 
integrity is required the as-left leakage rate will not exceed 0.75 
La. The combined leakage rate for containment isolation valves 
listed in Technical Specification Table 4.4-1 subject to gas or 
nitrogen pressurization testing, air lock testing, and portions of 
the sensitive leakage rate test which pertain to containment 
penetrations and double-gasketed seals shall be less than 0.6 La. 
The extensive operations and testing experience derived from 
industry show that risk to the general population is generally 
insensitive to changes in the allowable leakage rate. It has been 
determined that the allowable containment leakage can be increased 
by one to two orders of magnitude without significantly impacting 
the estimates of population dose in the event of an accident. 
Furthermore, the Indian Point Unit No. 2 ILRT test history provides 
substantial justification for the proposed changes.
    Test results demonstrate that IP2 [Indian Point 2] has a low 
leakage containment and that the proposed changes would not 
jeopardize the ability of the containment to maintain the leakage 
rate at or below the required limits. The proposed change to 
Technical Specification 4.1 Basis represent a minor editorial 
correction to the mathematical formula for minimum testing frequency 
which does not change the formula. Therefore, the probability and 
the consequence of a design basis accident are not being increased 
by the proposed changes.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Plant systems and components will not be operated in a different 
manner as a result of the proposed Technical Specification change. 
The proposed change permits a performance-based approach to 
determining the leakage-rate test frequency for the containment and 
containment penetrations (Type A, B, and C tests). There are no 
plant modifications, or changes in methods of operation. Therefore, 
the changes in testing intervals for the containment and containment 
penetrations have no affect on the probability of occurrence of a 
LOCA [loss-of-coolant-accident]. The Limiting Conditions for 
Operation are not being changed. Changing the as-found leakage-rate 
acceptance criterion to 1.0 La does not increase the probability or 
consequences of an accident. Changing the test interval for the 
containment and containment penetrations does not create any new 
accident precursors or methods of operation. The proposed change to 
Technical Specification 4.1 Basis represent a minor editorial 
correction to the mathematical formula for minimum testing frequency 
which does not change the formula. Therefore, the possibility for an 
accident of a different type than was previously evaluated in the 
safety analysis report is not created by the proposed Technical 
Specification.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    While the proposed changes do increase the probability for 
malfunction of equipment important to safety due to the longer 
intervals between leakage tests, it has been estimated that the 
longer test intervals will have an insignificant increase in the 
overall accident risk to the public. This increase has been reviewed 
and found to be acceptable by the NRC as documented in NUREG-1493 
and the recent rulemaking to 10 CFR 50 Appendix J. We also agree 
that this increase in accident risk is insignificant. Changing the 
as-found acceptance criterion to 1.0 La does not increase the 
consequences of an accident, since the accident analysis assume a 
leakage rate of La for design basis accidents. The as-left Type A 
test acceptance criterion remains at less than 0.75 La. Given that 
the Indian Point Unit No. 2 ILRT test history show no failures 
during plant life, the proposed changes should not lead to a 
significant probability of creating new leakage paths or increased 
leakage rates. The proposed change to Technical Specification 4.1 
Basis represent a minor editorial correction to the mathematical 
formula for minimum testing frequency which does not change the 
formula. Therefore, the accident analysis assumptions for design 
basis accidents are unaffected and the margin of safety is not 
decreased by the proposed Technical Specification change.
    Public Document Room location: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10610.

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

    Date of amendment request: January 18, 1996
    Description of amendment request: The proposed amendment would 
delete the requirement to perform inservice inspections of the primary 
coolant pump (PCP) flywheels.
    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 following evaluation supports the finding that operation of 
the facility in accordance with the proposed change to the Technical 
Specifications would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the Technical Specifications would delete 
the requirement to perform non-destructive examination of the upper 
flywheel on the PCPs. The fracture mechanics analyses conducted to 
support the change show that a preexisting crack sized just below 
detection level will not grow to the flaw size necessary to result 
in flywheel failure within the life of the plant. This analysis 
conservatively assumes minimum material properties, maximum flywheel 
accident speed, location of the flaw in the highest stress area and 
a number of startup/shutdown cycles eight times greater than 
expected. Since an existing flaw in the flywheel will not grow to 
the allowable flaw size under normal operating conditions or to the 
critical flaw size under LOCA [loss-of-coolant accident] conditions 
over the life of the plant, elimination of inservice inspection for 
such cracks during the plant's life will not involve a significant 
increase in the

[[Page 47977]]

probability of an accident previously considered.
    The proposed changes do not increase the amount of radioactive 
material available for release or modify any systems used for 
mitigation of such releases during accident conditions. Therefore, 
operation of the facility in accordance with the proposed change to 
the Technical Specifications would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed change to the Technical Specifications would not 
change the design, configuration, or method of operation of the 
plant and therefore, operation of the facility in accordance with 
the proposed change to the Technical Specifications would not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the Technical Specifications would not 
result in a significant reduction in the margin of safety. 
Significant conservatisms have been used for calculating the 
allowable flaw size, critical flaw size and crack growth rate in the 
PCP flywheels. These include minimum material properties, maximum 
flywheel accident speed, location of the postulated flaw in highest 
stress area and a number of startup/shutdown cycles eight times 
greater than expected. Since an existing flaw in the flywheel will 
not grow to the maximum allowable flaw size under normal operating 
conditions or to the critical flaw size under LOCA conditions over 
the life of the plant, elimination of inservice inspections for such 
cracks during the plant's life will not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: John Hannon

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

    Date of amendment request: August 8, 1996
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) of each unit to reference 
updated or recently approved methodologies used to calculate cycle-
specific limits contained in the Core Operating Limits Report (COLR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed changes will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative in nature, and do not 
affect any system, procedure, or manipulation of any equipment which 
could affect the probability or consequences of any accident.
    (2) The proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature, and cannot 
introduce any new failure mode or transient which could create any 
accident.
    (3) The proposed changes will not involve a significant 
reduction in a margin of safety.
    The proposed changes are administrative in nature, and will not 
affect any operating parameters or limits which could result in a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
proposed amendments involve no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 21, 1996
    Description of amendment request: The proposed amendments would 
revise the term ``lifting loads'' used in Technical Specification 
3.9.6b.2, Manipulator Crane, to ``lifting force.'' This revision will 
clarify that the static loads associated with the lifting tool, drive 
rod and control rod weights are not included in the lifting force 
limit.
    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:
    Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change is administrative in nature and does not 
represent any changes to the refueling process in the field. It more 
accurately describes the components for which the LCO's [Limiting 
Condition for Operation] protection is intended as well as giving a 
more accurate description of the auxiliary hoist's minimum capacity. 
It also broadens the domain of activities for which protective 
measures are taken by including drag load testing into monitored 
activities. At CNS [Catawba Nuclear Station], the auxiliary hoists 
and the manipulator cranes are rated at [greater than or equal to] 
3000 pounds and are surveillance tested to greater than 1000 pounds. 
This brackets the limit force lifting value change from 600 to 1000 
pounds in the amendment proposal.
    Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. This proposed administrative change reflects no changes in 
the refueling processes, or any systems, structures or components 
connected with the refueling process.
    Will the change involve a significant reduction in a margin of 
safety?
    No. The proposed administrative change has no impact on 
refueling processes, systems, structures or components, and does not 
result in any significant reduction in a margin of safety. The 
subject change only clarifies the original intent of the 
specification and more accurately describes the involved components, 
component capacities and the domain of activities for which measures 
are taken to protect the reactor internals.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: August 23, 1996 (TSCR 245)
    Description of amendment request: The amendment request proposes 
new pressure-temperature (P-T) limits up to

[[Page 47978]]

22, 27, and 32 effective full power years (EFPY). The new sets of P-T 
curves would be used beyond 17 EFPY in the future as the corresponding 
EFPY of operation is completed.
    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 determined that this change request with respect to P-T 
limits involves no significant hazards considerations in that 
operation of the Oyster Creek Plant in accordance with the proposed 
amendment, will not:
    1. Involve a significant increase in the probability of an 
accident because the new limits account for the increase in RT 
<INF>NDT, including statistical uncertainty, due to neutron 
irradiation of the reactor vessel as well as establishing initial RT 
<INF>NDT on the basis of current Code requirements, also including 
statistical uncertainty, in accordance with Reg. Guide 1.99, Rev. 2. 
The new P-T curves will assure that brittle fracture of the reactor 
vessel is prevented.
    2. Create the probability of a new or different kind of accident 
from any accident previously evaluated. These new limits are the 
result of the calculation methodology in Reg. Guide 1.99, Rev. 2 
[Radiation Embrittlement of Reactor Vessel Materials], as required 
by Generic Letter 88-11 [NRC Position on Radiation Embrittlement of 
Reactor Materials and its Impact on Plant Operations]. Primary 
system configuration and function remain unchanged.
    3. Involve a significant reduction in margin of safety because 
the bases for the margin of safety remain the same as current 
limits, i.e., ASME [American Society of Mechanical Engineers], Sect. 
XI, App. G for available fracture toughness and applied stress 
intensity, Reg. Guide 1.99, Rev. 2 for calculating applied stress 
intensity, Reg. Guide 1.99, Rev. 2 for calculating adjusted RT 
<INF>NDT and 10 CFR 50, App. G, for criticality conditions.
    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, NJ 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

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

    Date of amendment request: August 15, 1996
    Description of amendment request: The proposed amendment would 
modify the Clinton Power Station Technical Specifications to 
incorporate the revised Safety Limit Minimum Critical Power Ratio 
(SLMCPR) as calculated by General Electric (GE) for Cycle 7 operation. 
The need to change the SLMCPR resulted from the 10 CFR Part 21 
condition reported by GE in their letter to the NRC dated May 24, 1996.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (1) This change does not involve a significant increase in the 
probability or consequences of any accident previously evaluated. In 
lieu of utilizing a potentially nonconservative generic value, this 
change revises the SLMCPR to be appropriately conservative as it has 
been specifically calculated on a plant- and cycle-specific basis. 
Although the SLMCPR does not apply (i.e., is not assumed or required 
to be met) during any analyzed accident, the MCPR fuel cladding 
Safety Limit ensures that during normal operation and during 
anticipated operational occurrences (AOOs), at least 99.9% of the 
fuel rods in the core do not experience transition boiling. The 
revised value for the SLMCPR is determined using the same 
methodology as the previous SLMCPR with the exception that it 
utilizes plant specific conditions to determine the safety limit. 
The revised SLMCPR, therefore, accounts for actual expected power 
distributions in the Clinton Power Station (CPS) core as well as 
CPS-specific uncertainties. This provides a more conservative SLMCPR 
than the generic value used previously.
    The proposed change does not affect any of the parameters or 
conditions that contribute to initiation of any accidents previously 
evaluated. In addition, the proposed change does not affect the 
ability of any plant systems or equipment to operate as assumed in 
the safety analyses. The revised SLMCPR will continue to ensure that 
the fuel cladding integrity is not lost as a result of over-heating 
during normal plant operation or any AOO. As a result, the proposed 
change will not result in a significant increase in the consequences 
of any accident previously evaluated.
    (2) The proposed change does not involve any new modes or 
operation, any changes to setpoints, or any plant modifications. 
Further, the incorporation of a revised MCPR safety limit, which has 
been determined to be acceptable for CPS Cycle 7 operation, does not 
result in the creation of any new failure modes or potential 
precursors to an accident. Therefore, the proposed change does not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    (3) The proposed SLMCPR has been evaluated to ensure that during 
normal operation and during AOOs, at least 99.9% of the fuel rods in 
the core do not experience transition boiling. As noted above, the 
revised SLMCPR has been determined using the same methodology as 
used previously with the exception of using CPS Cycle 7 specific 
core and fuel design data. This change ensures that the margin of 
safety for fuel cladding integrity is maintained by providing a CPS 
specific MCPR safety limit as opposed to utilizing a potentially 
less conservative generic limit. Therefore, the implementation of 
the proposed change to the SLMCPR does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

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

    Date of amendment requests: May 8, 1996
    Description of amendment requests: The licensee proposes to revise 
improved Technical Specifications (TS) 3.9.4 and 3.9.5 to facilitate 
testing of low pressure safety injection system components and permit 
additional flexibility in scheduling maintenance on the shutdown 
cooling system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Limiting Conditions for Operation (LCO) in Technical 
Specifications (TSs) 3.9.4 and 3.9.5 define the operability 
requirements for the Shutdown Cooling (SDC) system during refueling 
operations (Mode 6) while the water level above the top of the 
reactor vessel

[[Page 47979]]

flange is at least 23 feet and less than 23 feet, respectively. The 
objective of these TSs is to ensure that 1) sufficient cooling is 
available to remove decay heat, 2) the water in the reactor vessel 
is maintained below 140 deg.F, and 3) sufficient coolant circulation 
is maintained in the reactor core to minimize boron stratification 
leading to a boron dilution incident.
    The proposed TS changes affect the current limits imposed while 
ensuring adherence to the bases of the TS. No plant modifications 
are being made. The reactor cavity water level limitations and SDC 
system required operating times are being changed based on plant 
specific calculations and the objectives of the TSs are being 
maintained.
    1) Reduce the water level where two loops of SDC are required 
from 23 feet to 20 feet above the reactor vessel flange,
    Prior to the approval of Unit 2 Amendment No. 127 and Unit 3 
Amendment No. 116, Technical Specification Bases Section 3/4.9.8 has 
stated that ``With the reactor vessel head removed and 23 feet of 
water above the reactor vessel flange, a large heat sink is 
available for core cooling, thus in the event of a failure of the 
operating shutdown cooling loop, adequate time is provided to 
initiate emergency procedures to cool the core.''
    In the Bases for the New Standard Technical Specifications, 
``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4 
it is stated that; ``The 23 ft level was selected because it 
corresponds to the 23 ft requirement established for fuel movement 
in LCO 3.9.6, ``Refueling Water Level.''
    Southern California Edison (Edison) calculations show that there 
is an insignificant difference in the time to boil due to the 3-foot 
change in required water level. Therefore, adequate water is still 
available to mitigate the consequences of losing SDC.
    2) Increase the time a required loop of the SDC system may be 
removed from service from up to 1 hour per 8-hour period to up to 2 
hours per 8-hour period, provided the upper guide structure has been 
removed from the reactor vessel,
    The proposed TS changes the time the SDC loop may be removed 
from operation from up to 1 hour per 8-hour period to up to 2 hours 
per 8-hour period, and allows removal of the SDC loop from operation 
for testing of the Low Pressure Safety Injection (LPSI) system 
components as well as for core alterations in the vicinity of the 
hot legs. The proposed TS change also imposes certain restrictions 
to ensure operating the SDC system in accordance with this proposed 
TS change is of no safety significance. These [r]estrictions are 
discussed separately below.
    Specifically stating that the upper guide structure will be 
removed assures that natural heat transfer is not impeded.
    When securing the only operating loop of the SDC system the 
maximum Reactor Coolant System (RCS) temperature is maintained [less 
than or equal to] 140 deg.F. The initial conditions and heatup rate 
are selected such that the RCS temperature remains [less than or 
equal to] 140 deg.F during the test. Therefore, there is ample 
margin to boiling. Typical initial temperatures are less than 
100 deg.F.
    The water being injected by the LPSI system test is cool water 
from the Refueling Water Storage Tank (RWST) and will increase the 
reactor cavity water level by several inches, providing more cool 
water to the heat sink. The two hours is sufficient time to align 
the system to test, perform the test, and restore the loop of SDC to 
operation prior to exceeding 140 deg.F.
    No operations are permitted that would cause a reduction of the 
RCS boron concentration. This minimizes the probability of an 
inadvertent boron dilution event. The use of adequately borated 
water for injection into the RCS during the test provides assurance 
that the test itself cannot lead to a boron dilution event. When the 
SDC system is operating, the minimum SDC flow rate of 2200 gpm 
imposed by Surveillance Requirements SR 3.9.4.1 and SR 3.9.5.1 is 
sufficient to ensure complete mixing of the boron within the RCS.
    Securing SDC flow is only allowed when the reactor cavity water 
level is maintained greater than or equal to 20 feet above the 
reactor vessel flange. This level ensures an adequate heat sink to 
perform the LPSI pump suction header check valve test.
    3) Allow for running 1 loop of shutdown cooling with additional 
requirements when the water level is less than 20 feet but greater 
than or equal to 12 feet above the reactor vessel flange,
    4) Add an action to be taken when operating 1 loop of SDC with 
less than 20 feet of water above the reactor vessel flange when the 
specified requirements are not met,
    In the event of a loss of SDC the time to boil is reduced from 
approximately 4.0 hours when the water level is 23 feet above the 
reactor vessel flange to approximately 2.3 hours at 12 feet, 
assuming the reactor has only been shutdown for 6 days. However, 
this is ample time to close containment (less than 1 hour) and to 
restore SDC or initiate alternative cooling (e.g., add water to the 
cavity (approximately 1 hour)). The reactor pressure vessel flange 
is approximately 11' above the top of the fuel. Therefore, the water 
level will be a minimum of 23' above the fuel, which still maintains 
a large volume of water to provide a heat sink.
    Requiring the reactor to be shutdown for at least 6 days to have 
only one loop of SDC operable when the reactor cavity level is 
between 20 feet and 12 feet above the reactor vessel flange ensures 
that the time to boil is greater than twice the time it would take 
to establish containment closure and to commence reactor cavity fill 
with the required standby equipment.
    One loop of SDC operating with a containment spray pump allows 
for the high capacity LPSI pump to be the main standby pump capable 
of filling the reactor cavity to at least 20 feet above the reactor 
pressure vessel flange in the event SDC is lost. The high pressure 
safety injection pump will also be maintained OPERABLE to increase 
the water level if needed. In support of this contingency the RWST 
will be required to contain the volume of water needed to raised 
[raise] the level to 20 feet above the reactor pressure vessel 
flange. As discussed above, the reactor cavity can be filled at a 
rate of approximately 4.0 inches per minute with the LPSI pump.
    If operating one loop of the SDC system with less than 20 feet 
of water above the reactor vessel flange and any of the required 
conditions are not met, requiring immediate action to establish 
greater than or equal to 20 feet of water above the reactor vessel 
flange ensures no time is wasted trying to restore the required 
condition not met. By taking action to restore the level to 20 feet 
above the reactor vessel flange the plant will be placed in TS 
3.9.4, which only requires one loop of SDC to be operable. 
Additionally, the core will not heat up while the water level in the 
reactor cavity is being raised with cool water from the RWST. This 
will provide additional time to either restore the one loop of SDC 
or take other actions to provide core cooling as required by TS 
3.9.4.
    A Probabilistic Risk Assessment (PRA), with a) one loop of the 
SDC system operable with the reactor cavity water level greater than 
or equal to 12 feet above the reactor vessel flange, and b) one loop 
of the SDC system operable with the reactor cavity water level 
greater than or equal to 20 feet above the reactor vessel flange, 
showed that the operations in accordance with the proposed TS would 
not significantly increase the probabilities of inventory boiling 
and core damage.
    5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that 
was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit 
3 Amendment No. 116.
    This is an editorial change.
    Therefore, proposed changes 1 through 5 do not involve a 
significant increase in the probability or consequences of an 
accident.
    2. The proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    1) Reduce the water level where two loops of SDC are required 
from 23 feet to 20 feet above the reactor vessel flange,
    2) Increase the time a required loop of the SDC system may be 
removed from service from up to 1 hour per 8-hour period to up to 2 
hours per 8-hour period, provided the upper guide structure has been 
removed from the reactor vessel,
    3) Allow for running 1 loop of shutdown cooling with additional 
requirements when the water level is less than 20 feet but greater 
than or equal to 12 feet above the reactor vessel flange,
    4) Add an action to be taken when operating 1 loop of SDC with less 
than 20 feet of water above the reactor vessel flange when the 
specified requirements are not met,
    The Limiting Conditions for Operation (LCO) in Technical 
Specifications (TSs) 3.9.4 and 3.9.5 define the operability 
requirements for the SDC system during refueling operations (Mode 6) 
while the water level above the top of the reactor vessel flange is 
at least 23 feet and less than 23 feet, respectively. The objective 
of the proposed TS changes is to ensure that the intent of the Bases 
is maintained. [i.e., 1) sufficient cooling is available to remove 
decay heat, 2) water in the reactor vessel is maintained below 
140 deg.F, and 3) sufficient coolant

[[Page 47980]]

circulation is maintained in the reactor core to minimize boron 
stratification leading to a boron dilution incident.]
    The proposed TS changes affect the current limits imposed while 
ensuring adherence to the bases of the TS. No plant modifications 
are being made. The reactor cavity water level limitations and SDC 
system required operating times are being changed based on plant 
specific calculations, and the objective of the TSs are being 
maintained. The added requirements and action statement facilitate 
safe operation.
    5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that 
was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit 3 
Amendment No. 116.
    This is an editorial change.
    Therefore, the operation of the facility in accordance with 
proposed changes 1 through 5 does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    Limiting Conditions for Operation (LCO) in TSs 3.9.4 and 3.9.5 
define the operability requirements for the SDC system during 
refueling operations (Mode 6) while the water level above the top of 
the reactor vessel flange is at least 23 feet and less than 23 feet, 
respectively. The objectives of these TSs are to ensure that 1) 
sufficient cooling is available to remove decay heat, 2) the water 
in the reactor vessel is maintained below 140 deg.F, and 3) 
sufficient coolant circulation is maintained in the reactor core to 
minimize boron stratification leading to a boron dilution incident.
    1) Reduce the water level where two loops of SDC are required 
from 23 feet to 20 feet above the reactor vessel flange,
    Prior to the approval of Unit 2 Amendment No. 127 and Unit 3 
Amendment No. 116, Technical Specification Bases Section 3/4.9.8 has 
stated that ``With the reactor vessel head removed and 23 feet of 
water above the reactor vessel flange, a large heat sink is 
available for core cooling, thus in the event of a failure of the 
operating shutdown cooling loop, adequate time is provided to 
initiate emergency procedures to cool the core.''
    In the Bases for the New Standard Technical Specifications, 
NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4 it 
is stated that ``The 23 ft level was selected because it corresponds 
to the 23 ft requirement established for fuel movement in LCO 3.9.6, 
``Refueling Water Level.''
    Edison calculations show that there is a minimal difference in 
the time to boil due to the 3-foot change in required water level. 
Therefore, the margin of safety has not been significantly reduced.
    2) Increase the time a required loop of the SDC system may be 
removed from service from up to 1 hour per 8-hour period to up to 2 
hours per 8-hour period, provided the upper guide structure has been 
removed from the reactor vessel,
    The proposed TS changes the time the SDC loop may be removed 
from operation from up to 1 hour per 8-hour period to up to 2 hours 
per 8-hour period, and allows removal of the SDC loop from operation 
for testing of the LPSI system components as well as for core 
alterations in the vicinity of the hot legs. The proposed TS change 
also imposes certain restrictions to ensure operating the SDC system 
in accordance with this proposed TS change is of no safety 
significance. These restrictions are discussed separately below.
    Specifically stating that the upper guide structure will be 
removed assures that natural heat transfer is not impeded.
    When securing the only operating loop of the SDC system, the 
maximum RCS temperature is maintained [less than or equal to] 
140 deg.F. The initial conditions and heatup rate are selected such 
that RCS temperature remains [less than or equal to] 140 deg.F 
during the test. Therefore, there is ample margin to boiling. 
Typical initial temperatures are less than 100 deg.F.
    The water being injected by the LPSI system test is cool borated 
water from the RWST and will increase the level of the reactor 
cavity by several inches. The two hours is sufficient time to align 
the system to test, perform the test, and restore the loop of SDC to 
operation prior to exceeding 140 deg.F.
    No operations are permitted that would cause a reduction of the 
RCS boron concentration. This minimizes the probability of an 
inadvertent boron dilution event. The use of adequately borated 
water for injection into the RCS during the test provides assurance 
that the test itself cannot lead to a boron dilution event. When the 
SDC system is operating, the minimum SDC flow rate of 2200 gpm is 
sufficient to ensure complete mixing of the boron within the RCS.
    Securing SDC flow is only allowed when the reactor cavity water 
level is maintained greater than or equal to 20 feet above the 
reactor vessel flange. This level ensures an adequate heat sink to 
perform the LPSI pump suction header check valve test.
    The added requirements and the nature of the test provide 
assurances that the water temperature will be maintained less than 
140 deg.F and that boron stratification is prevented.
    3) Allow for running 1 loop of shutdown cooling with additional 
requirements when the water level is less than 20 feet but greater 
than or equal to 12 feet above the reactor vessel flange,
    4) Add an action to be taken when operating 1 loop of SDC with less 
than 20 feet of water above the reactor vessel flange when the 
specified requirements are not met,
    In the event of a loss of SDC, the time to boil is reduced from 
approximately 4.0 hours when the water level is 23 feet above the 
reactor vessel flange to approximately 2.3 hours at 12 feet, when 
the reactor has only been shutdown for 6 days. However, this is 
ample time to close containment (less than 1 hour), and to restore 
SDC or initiate alternative cooling (e.g., add water to the cavity 
(approximately 1 hour)).
    Requiring the reactor to be shutdown for at least 6 days to have 
only one loop of SDC operable when the reactor cavity level is 
between 20 feet and 12 feet above the reactor vessel flange ensures 
that the time to boil is greater than twice the time it would take 
us to establish containment closure and to commence reactor cavity 
fill with the required standby equipment.
    One loop of SDC operating with a containment spray pump allows 
for the high capacity LPSI pump to be the main standby pump capable 
of filling the reactor cavity to at least 20 feet above the reactor 
pressure vessel flange in the event SDC is lost. The high pressure 
safety injection pump will also be maintained OPERABLE to increase 
the water level if needed. In support of this contingency the RWST 
will be required to contain the volume of water needed to raised 
[raise] the level to 20 feet above the reactor pressure vessel 
flange. As discussed above, the reactor cavity can be filled at a 
rate of approximately 4.0 inches per minute with the LPSI pump.
    If operating one loop of the SDC system with less than 20 feet of 
water above the reactor vessel flange and any of the required 
conditions are not met, requiring immediate action to establish greater 
than or equal to 20 feet of water above the reactor vessel flange 
ensures no time is wasted trying to restore the required condition not 
met. By taking action to restore the level to 20 feet above the reactor 
vessel flange the plant will be placed in TS 3.9.4, which only requires 
one loop of SDC to be operable. Additionally, the core will not heat up 
while the reactor cavity water level is being raised with cool water 
from the RWST. This will provide additional time to either restore the 
one loop of SDC or take other actions to provide core cooling as 
required by TS 3.9.4.
    A PRA showed that operations in accordance with the proposed TS 
did not significantly increase the probabilities of inventory 
boiling and core damage.
    5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that 
was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit 3 
Amendment No. 116.
    This is an editorial change.
    Therefore, operation of the facility in accordance with proposed 
changes 1 through 5 do not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration. 
Temporary
    Local Public Document Room location: Science Library, University of 
California, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

[[Page 47981]]

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

    Date of amendment requests: May 9, 1996, as supplemented by letter 
dated June 27, 1996.
    Description of amendment requests: The licensee proposes to add a 
requirement to maintain a Barrier Control Program to Section 5 of the 
improved Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
1. The proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change will allow a passive support system, plant 
barriers, to be taken out of service for a specific allowed outage 
time. Since the allowed outage times are to limit the average annual 
cumulative increase in fuel damage risk to less than 1.0E-6, there 
will not be a significant increase in either the probability or 
consequences of any accident previously evaluated. Additionally, the 
proposed change will allow barrier impairments if allowed by a 10 
CFR 50.59 evaluation and also if the equipment is declared 
inoperable or is not needed. Since these two conditions are already 
a part of the San Onofre Units 2 and 3 Licensing Basis, there will 
be no change in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Barriers have been analyzed for specific hazards. The nature of 
these hazards will not change due to this amendment, and therefore 
no new or different kind of accident will be created from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Since allowing barrier impairments in accordance with 10 CFR 
50.59 or declaring affected equipment inoperable is part of the 
SONGS Units 2 and 3 Licensing Basis, there will be no reduction in 
the margin of safety from these two criteria.
    Allowing allowed outage times for barrier impairments does not 
have a significant effect on a margin of safety because the average 
annual cumulative increase in fuel damage risk is limited to less 
than 1.0E-6/yr. This small increase is about 3% of the San Onofre 
Units 2 and 3 core damage risk as reported in the Individual Plant 
Examination (IPE).
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration. 
Temporary
    Local Public Document Room location:  Science Library, University 
of California, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: May 29, 1996
    Description of amendment requests: The licensee proposes to revise 
the acceptance criteria for the Agastat time delay relays used in the 
engineered safety features (ESF) load sequencer in Surveillance 
Requirement (SR) 3.8.1.18, ``A.C. Sources - Operating'' of Technical 
Specification (TS) 3.8.1, ``A.C. Sources - Operating.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change will not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The proposed change would expand the current surveillance 
acceptance criteria to more accurately reflect the characteristics 
of the installed plant equipment. The diesel generators (DG's) have 
sufficient capacity to maintain adequate voltage and frequency 
during load sequencing with the expanded tolerance. The overall 
Engineered Safety Features (ESF) response times in the Technical 
Specifications and safety analyses are maintained even though the 
timer tolerance is increased, therefore, the consequences of any 
accident previously evaluated are not increased. The DG load 
sequence timers are not of themselves a credible initiator of any 
accident, so the probability of an accident has not been increased. 
The timers will function acceptably to support the equipment needed 
for accident mitigation, so the consequences of an accident are not 
increased. Therefore, the probability or consequences of any 
accident previously evaluated is not increased.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This amendment request does not involve any change to plant 
equipment or operation. In the event of a loss of preferred power, 
the ESF electrical loads are automatically connected to the DG's in 
sufficient time to provide for safe reactor shutdown and to mitigate 
the consequences of a Design Basis Accident (DBA) such as a loss of 
coolant accident (LOCA). Increasing the timer tolerance will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change will not involve a significant reduction 
in a margin of safety.
    This amendment does not change the manner in which safety 
limits, limiting safety settings, or limiting conditions for 
operations are determined. The actual response times have not been 
altered by this amendment, therefore, operations will not be 
affected. Accordingly, this amendment will not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration. 
Temporary
    Local Public Document Room location: Science Library, University of 
California, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendment requests: May 30, 1996
    Description of amendment requests: The licensee proposes to revise 
Surveillance Requirements (SR) 3.6.1.1, 3.6.2.1, and 3.6.3.6, of the 
improved Technical Specifications. The proposed change will allow 
implementation of the recently approved Option B to 10 CFR Part 50, 
Appendix J. This new rule allows for a performance-based option for 
determining the test frequency for containment leakage rate testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Since the interval between containment leakage rate tests is not 
related in any way to conditions which cause accidents, and plant 
structures, systems, and components

[[Page 47982]]

will not be operated in a different manner as a result of the 
proposed Technical Specification (TS) change, the proposed changes 
will not increase the probability of an accident previously 
evaluated.
    Containment leakage may result from accidents which are 
evaluated in the Updated Final Safety Analysis Report. The proposed 
TS changes may result in an acceptably small increase in post-
accident containment leakage. Using a statistical approach, NUREG-
1493 determined that the increase in hypothetical dose to the public 
resulting from extending the testing interval is extremely small. 
NUREG-1493 concluded that such small hypothetical dose increases to 
the public are justifiable due to the real reduction in occupational 
exposure resulting from interval extension. Therefore, the proposed 
change does not significantly increase the consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change only incorporates the performance based 
approach for containment leak rate testing authorized in the new 
Option B to Appendix J of 10 CFR Part 50. The interval extensions 
allowed, through this approach, do not have the potential for 
creating the possibility of new or different kinds of accidents from 
those previously evaluated because plant structures, systems, and 
components will not be operated in a different manner as a result of 
the TS change and, therefore, will not introduce any new or 
different failure modes or initiators. Therefore the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed Technical Specification does not alter the 
allowable containment leakage rate. The proposed change replaces the 
current, prescriptive testing requirements with a new performance 
based approach for establishing the testing intervals. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration. 
Temporary
    Local Public Document Room location: Science Library, University of 
California, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

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

    Date of amendments request: August 23, 1996
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications to allow installation of laser 
welded elevated tubesheet sleeves in Farley, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the Farley Nuclear Plant Units 1 and 2 steam 
generators in accordance with the proposed license amendment does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The installation of elevated tubesheet laser welded sleeves as 
described below, can be used to repair degraded tubes by returning 
the condition of the tubes to their original design condition (for 
tube integrity, stress and fatigue considerations, and leaktightness 
during all plant conditions). Tube bundle overall structural and 
leakage integrity will be increased with the installation of the 
laser welded sleeves. The performance history of Westinghouse 
sleeves has shown that, to date, no domestic laser welded sleeves 
have been removed from service due to corrosion degradation of the 
sleeve or parent tube in the joint area.
    Any hypothetical sleeve failure is bounded by the consequences 
of a postulated steam generator tube rupture event. The use of 
elevated tubesheet laser welded sleeves will not increase the amount 
of primary-to-secondary leakage anticipated during a postulated 
steam linebreak and other analyzed accidents. Leak rate tests show 
only negligible primary-to-secondary leakage through the non-welded 
elevated tubesheet sleeve lower joints during normal or accident 
conditions such that any consequences are insignificant with regard 
to offsite doses. Sleeve installation will result in an increase in 
resistance to primary coolant flow through the tube. Depending on 
the assumed steam generator tube rupture location, the primary 
coolant flow through the ruptured tube is reduced by the influence 
of sleeves installed below the break location, thereby reducing the 
consequences to the public due to a steam generator tube rupture 
event. Steam generator tube sleeving has as a basis that the 
analyzed steam generator tube plugging level and associated minimum 
measured flow rate, is not exceeded. Therefore, primary coolant flow 
area assumptions in the accident analyses are not affected and any 
consequences of a postulated loss of coolant accident would not be 
increased.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Installation of elevated tubesheet laser welded sleeves will 
increase the leaktightness of the tube bundle in addition to 
enhancing overall steam generator tube bundle integrity by isolating 
localized tube wall degradation. Isolation of the tube degradation 
is provided by attachment between the tube and sleeve at each end of 
the sleeve. Following the installation of the sleeves, steam 
generator tube integrity is restored to its original design bases.
    Testing has shown that once installed, there is no mechanism for 
the sleeves to affect any portion of the steam generator other than 
the tubes in which they are installed. No other system or component 
connecting with the steam generator is adversely affected by the 
operation of the steam generator following installation of laser 
welded tube sleeves.
    Structural analyses of the tube, sleeve and sleeve joints show 
the stress limits defined in the ASME [American Society of 
Mechanical Engineers] Code are not exceeded during all plant 
conditions. The effect of any hypothetical failure of the sleeve 
would be bounded by existing tube rupture analyses. No increase in 
leakage is anticipated during a postulated steam line break event. 
Therefore, operation of the steam generators following installation 
of elevated tubesheet laser welded sleeves in the tubes of the 
Farley steam generators will not result in an accident previously 
not analyzed in the FSAR [Final Safety Analysis Report].
    Therefore, SNC [Southern Nuclear Operating Company] concludes 
that the proposed license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The margin of safety with respect to maintenance of the 
integrity of the tube bundle is provided, in part, by the safety 
factors included in the ASME Code, and is not reduced. 
Nondestructive examination of the sleeve and non-sleeved tube length 
still can be performed; therefore, the recommendations of Regulatory 
Guide 1.83, Revision 1 can be implemented. The installation process 
of the elevated tubesheet laser welded sleeves has been shown to 
provide an essentially leaktight bond between the sleeve and the 
tube during all plant conditions, and, as such, would not 
significantly contribute to the radiological consequences of a 
postulated steam line break event. Any combination of sleeving and 
plugging utilized at Farley Units 1 and 2 up to the level that 
analyzed minimum measured reactor coolant flow rate is maintained 
per Technical Specification requirements, will be bounded by the 
accident analyses supporting the analyzed flow level.
    Therefore, SNC, concludes that the proposed change does not 
result in a significant reduction in a loss of margin with respect 
to plant safety as defined in the Final Safety Analysis Report or 
the bases of the Farley technical specifications.
    Based on the preceding analysis, it is concluded that operation 
of the Farley

[[Page 47983]]

Nuclear Plant steam generators in accordance with the proposed 
amendment does not involve a significant hazards consideration as 
defined in 10 CFR 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: June 29, 1996 (TS 5.2.2.f)
    Description of amendment request: The proposed amendment would 
revise the Watts Bar (WBN) Unit 1 Technical Specification (TS) 
requirements to delete the first sentence of TS Section 5.2.2.f which 
reads, ``The Operations Manager shall hold or have held an SRO [Senior 
Reactor Operator] license on a similar unit.'' The remaining sentence 
of this section is being revised to indicate that the Operations 
Superintendent will hold an SRO license for WBN Unit 1. This change is 
consistent with the Tennessee Valley Authority's (TVA) commitment to 
ANSI N18.1-1971 regarding the qualification of this position and is 
consistent with the Standard TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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.
    Operation of the plant in accordance with the proposed amendment 
will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    As explained in the June 29, 1996 submittal, the proposed change is 
considered to be administrative in nature. The proposed change affects 
an administrative control, which was based on the guidance of ANSI 
N18.-1971. ANSI N18.1-1971 recommended that the Operations Manager hold 
an SRO license. The ANSI N18.1-1971 Standard defines the positions of 
Plant Manager, Operations Manager, Supervisors and Operators. A 
subsequent update of this standard, ANSI/ANS 3.1-1987, also defines the 
position of Operations Middle Manager. The correlating named positions 
in the TVA management structure at WBN are: WBN Operations Manager 
correlates to ANSI Plant Manager, WBN Operations Superintendent 
correlates to ANSI Operations Manager or Operations Middle Manager, WBN 
Shift Operations Supervisor correlates to ANSI Shift Supervisor, and 
WBN Senior and Licensed Operators correlate to ANSI operators. The 
guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends that ``If the 
Operations Manager does not hold an NRC License, then the Operations 
Middle Manager shall hold an NRC Senior Operator's License. This would 
be consistent with TVA's proposal that the WBN Operations 
Superintendent (ANSI Operations Middle Manager) continue to be required 
to maintain an SRO license.
    The proposed change does not alter the design of any system, 
structure, or component, nor does it change the way plant systems are 
operated. It does not reduce the knowledge, qualifications, or skills 
of licensed operators. The control room operators will continue to be 
supervised by the licensed Shift Supervisors and the first level of 
off-shift WBN managemet directing the activities of licensed operators 
will continue to hold an SRO license. In summary, the proposed change 
does not affect the ability of the Operations Superintendent to provide 
the plant oversight required of his position. Thus, it does not involve 
a significant increase in the probability or consequence of an accident 
previously evaluated.
    (2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed change to TS 5.2.2.f does not affect the design or 
function of any plant system, structure, or component, nor does it 
change the way plant systems are operated. It does not affect the 
performance of NRC licensed operators. Operation of the plant will 
continue to be supervised by personnnel who hold an NRC SRO license. 
Based on the above, the proposed change does not create the possibility 
of a new or different kind of accident from any previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed change involves an administrative control. The 
proposed change does not reduce the level of knowledge or experience 
required of an individual who fills the Operations Superintendent 
position. The control room operators will continue to be supervised by 
personnel who hold an SRO license. Thus, the proposed change does not 
ivnolve a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: November 17, 1995, as supplemented July 
29, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 15.6.3, ``Facility Staff 
Qualifications.'' The title of the responsible health physicist would 
be changed, and a requirement for this individual to be a supervisor 
would be added.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of an accident previous evaluated.
    The proposed changes separate the qualifications requirements of 
the Technical Specifications from the Health Physics Manager, while 
requiring that the same qualifications be fulfilled by a designated 
Health Physicist position within the organization. This change 
maintains the present knowledge requirements of the PBNP [Point 
Beach Nuclear Plant] staff. The personnel holding the health physics 
qualifications are not considered in the probability of any 
accident. By ensuring the appropriate expertise remains on the staff 
to advise management on issues related to radiological safety, 
appropriate action is assured during analyzed events to assess and 
mitigate the radiological consequences. Therefore, this change does 
not affect the probability or consequences of any accident 
previously evaluated.

[[Page 47984]]

    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a new or different kind 
of accident from any accident previously evaluated.
    The proposed change separates the Health Physics Manager 
qualifications from the position while maintaining the requirements 
for that expertise to be maintained within the organization. This is 
an administrative change only and does not affect any plant 
structures, systems or components. Therefore, a new or different 
kind of accident from any accident previously evaluated cannot 
result.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant reduction 
in a margin of safety.
    The proposed changes are administrative only. The required 
levels of expertise and experience will be maintained within the 
Health Physics organization. Therefore, there is no 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: 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: Gail H. Marcus

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: January 5, 1996, as supplemented July 
12, 1996
    Description of amendment request: The proposed amendment would 
revise the requirements of technical specification 3.1.9.3 to permit a 
filled refueling cavity to serve as a back-up means of decay heat 
removal.
    Date of individual notice in the Federal Register: August 28, 1996 
(61 FR 44348)
    Expiration date of individual notice: September 27, 1996
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendments: March 15, 1995, as supplemented 
June 29, 1995, May 1, 1996 and May 15, 1996.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) Section 6.0, ``Administrative Controls'' 
to be consistent with the guidance provided in the Improved Standard 
Technical Specifications (STSs) for Combustion Engineering Plants. 
Additionally, the amendments (a) allow the Shift Technical Advisory to 
perform dual roles, (b) establishes a TS Bases Control Program, (c) 
provides for a reduction in the reporting requirements, and (d) 
provides an option for estimating occupational doses.
    Date of issuance: August 26, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 216 and 193
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42598) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated August 26, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County and Northeast Nuclear Energy Company, 
et al., Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear 
Power Station, Units 1, 2, and 3, New London County, Connecticut

    Date of application for amendments: November 22, 1995
    Brief description of amendments: The amendments replace the title-
specific designation of members representing specific functional areas 
on the Plant Operating Review Committee (PORC) for the Haddam Neck 
Plant and Millstone Units 1, 2, and 3 with a functional area-specific 
designation that stipulates membership qualification and experience 
requirements. The amendments also clarify the composition of the Site 
Operations Review Committee (SORC) at Millstone.
    Date of issuance: July 16, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.

[[Page 47985]]

    Amendment Nos.: 190, 95, 200, 130
    Facility Operating License Nos. DPR-61, DPR-21, DPR-65, AND NPF-49: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7549) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 16, 1996 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and 
the Learning Resources Center, Three Rivers Community-Technical 
College, 574 New London Turnpike, Norwich, Connecticut 06360, and 
Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, 
Connecticut 06385, for Millstone 1, 2, and 3.

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: June 6, 1996; supplemented 
August 1, 1996
    Brief description of amendments: The amendments revise the 
Technical Specification requirements related to testing of the Low 
Pressure Service Water pumps and valves, LPSW-4 and LPSW-5, to reflect 
a design change to remove the Engineered Safeguards signal from the 
valves.
    Date of Issuance: August 19, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 217, 217, 214
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37298) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 19, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina

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

    Date of application for amendment: May 31, 1996, as supplemented by 
letter dated May 2, 1996
    Brief description of amendment: The amendment revised the schedule 
for withdrawing capsules with reactor vessel material specimens in 
accordance with the reactor vessel material surveillance program for 
the Grand Gulf Nuclear Station, Unit 1 and Section III.B.3 of Appendix 
H, ``Reactor Vessel Material Surveillance Program Requirements,'' of 10 
CFR Part 50.
    Date of issuance: August 21, 1996
    Effective date: August 21, 1996
    Amendment No: 127
    Facility Operating License No. NPF-29: Amendment revises the 
license.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31179) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 21, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

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

    Date of application for amendments: May 17, 1995, as supplemented 
July 15, 1996.
    Brief description of amendments: These amendments improve 
consistency between the Technical Specifications (TS) and the improved 
Combustion Engineering Standard Technical Specifications (STS) and 
resolve other inconsistencies in the TS.
    Date of Issuance: August 14, 1996
    Effective Date: August 14, 1996
    Amendment Nos.: 146 and 85
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32363). The July 15, 1996, letter made a minor change to the proposed 
definition of core alteration which made it more closely match the 
wording in the STS and did not change the scope of the May 17, 1995, 
application and initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 14, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendments: August 16, 1996
    Brief description of amendments: Relocates selected Technical 
Specifications (TS) related to instrumentation to the Updated Final 
Safety Analysis Report, in accordance with the Commissions Final Policy 
Statement on TS Improvement for Nuclear Power Reactors (58 FR 39132, 
July 22, 1993). Also relocates review requirements related to the 
Emergency Plan and the Security Plan from the TS to the respective 
plans.
    Date of Issuance: August 20, 1996
    Effective Date: August 20, 1996
    Amendment Nos.: 147 and 86
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49938) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 20, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendments: June 17, 1996
    Brief description of amendments: The amendments revise Technical 
Specification 5.3.1, Fuel Assemblies, to remove the restriction on the 
number of fuel rods clad with ZIRLO<SUP>TM that can be loaded into the 
core.
    Date of issuance: August 19, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 94, 72
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37299)

[[Page 47986]]

The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 19, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

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

    Date of application for amendments: June 17, 1996
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.8.1, A.C. Sources, and its associated Bases, by 
changing Surveillance Requirement 4.8.1.1.2.j(2) to limit the 10-year 
pressure test of certain portions of the diesel fuel oil system to the 
isolable portions of the fuel oil piping.
    Date of issuance: August 28, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 95 and 73
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37300) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 28, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

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

    Date of application for amendment: March 15, 1996, as supplemented 
July 18, 1996.
    Brief description of amendment: The amendment revised TS 4.6.2.1 
``Containment Systems - Depressurization Systems - Suppression Pool'' 
to extend the time interval for performing the containment drywell-to-
suppression chamber bypass leakage tests consistent with schedules for 
containment integrated leak rate testing under Option B to 10 CFR Part 
50, Appendix J.
    Date of issuance: August 27, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 75
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20851) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 27, 1996 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: March 20, 1996
    Brief description of amendment: The amendment revises Technical 
Specification 3/4.3.1 ``Reactor Protection System Instrumentation'' to 
modify operability requirements for the Average Power Range Monitor for 
operational conditions 3, 4, and 5.
    Date of issuance: August 28, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 76
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20852) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 28, 1996 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: April 25, 1996
    Brief description of amendment: The amendment modifies the 
calibration requirement for the source range monitors and intermediate 
range monitors by noting that the sensors are excluded.
    Date of issuance: August 19, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 96
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31183) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 19, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

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

    Date of application for amendments: November 14, 1994, as 
supplemented by letters dated December 7, 1995, February 2, 1996, May 
28, 1996, and July 30, 1996.
    Brief description of amendments: The amendment revised the combined 
Technical Specifications (TS) for the Diablo Canyon Nuclear Power 
Plant, Unit Nos. 1 and 2, for the slave relay test frequency from 
quarterly (Q) to refueling (R). The request also removed table notation 
4 from Table 4.3-2. The associated Bases were revised.
    Date of issuance: August 19, 1996
    Effective date: August 19, 1996, to be implemented within 30 days 
of date of issuance.
    Amendment Nos.: Unit 1 - 115; Unit 2 - 113
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62495). The supplemental letters provided additional clarifying 
information and did not change the original no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated August 19, 1996. 
No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

[[Page 47987]]

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

    Date of application for amendment: June 13, 1996, as supplemented 
by letter dated August 7, 1996.
    Brief description of amendment: This amendment will permit a one 
time performance of TS surveillance requirement 3.3.1.1.12 for the 
Average Power Range Monitor Flow Biased High Scram function with a 
delayed entry into associated TS Conditions and Required Actions for up 
to six hours provided core flow is maintained at or above eighty-two 
percent. This change is in effect until the end of refueling outage 
2R11.
    Date of issuance: August 16, 1996
    Effective date: Unit 2, as of the date of issuance, to be 
implemented within 30 days.
    Amendment No.: 216
    Facility Operating License No. DPR-44: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34895) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 16, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

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

    Date of applications for amendment: June 15, September 15, October 
25, and November 30, 1995.
    Brief description of amendment: The amendments change the Technical 
Specifications regarding the Control Rod System, the Auxiliary 
Electrical Systems, the Containment Systems and the Standby Liquid 
Control System to reflect changes to the length of the operating cycle 
of 24 months.
    Date of issuance: August 16, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 232
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47623), January 22, 1996 (61 FR 1633, 61 FR 1634, 61 FR 1635) The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 16, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Dated at Rockville, Maryland, this 4th day of September 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 96-23032 Filed 9-10-96; 8:45 am]
BILLING CODE 7590-01-F