[Federal Register Volume 61, Number 90 (Wednesday, May 8, 1996)]
[Notices]
[Pages 20842-20865]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-11295]



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NUCLEAR REGULATORY COMMISSION

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 13, 1996, through April 26, 1996. The 
last biweekly notice was published on April 24, 1996 (61 FR 18162).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, 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 June 7, 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

[[Page 20843]]

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, 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: March 28, 1996.
    Description of amendments request: Pursuant to 10 CFR 50.90, the 
Baltimore Gas and Electric Company (BGE) hereby requests an amendment 
to Operating License Nos. DPR-53 and DPR-69 to reduce the moderator 
temperature coefficient (MTC) limit shown on Technical Specification 
Figure 3.1.1-1. This proposed change is necessary to support changes in 
the safety analyses made to accommodate a larger number of plugged 
steam generator (SG) tubes for future operating cycles. The proposed 
limit will be more restrictive than the existing limit to match the 
analytical assumptions. In addition, the licensee provided information 
to clarify the relationship of the MTC to an Anticipated Transient 
Without Scram event in its licensing basis.
    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 safety analyses for the current fuel cycles assume 500 tubes 
per steam generator (SG) are plugged and the maximum beginning-of-
cycle moderator temperature coefficient (MTC) is assumed to follow 
the curve in Technical Specification Figure 3.1.1.-1. For the fuel 
cycle to be installed in Unit 1 in spring 1996, Baltimore Gas and 
Electric Company (BGE) assumes in the analyses that more SG tubes 
are plugged than the current limit, and it is necessary to credit a 
more restrictive (less positive) limit on the maximum positive MTC 
to mitigate the

[[Page 20844]]

Reactor Coolant System pressure and temperature increase analyzed 
for these events. Therefore, we are proposing a change to the 
allowable positive MTC limits shown on Technical Specification 
Figure 3.1.1-1. The proposed limit will be more restrictive than the 
existing limit to match the analytical assumptions. Since the safety 
analyses supporting an increase in the number of plugged SG tubes 
are applicable to both Units 1 and 2, BGE is requesting this change 
for both Units.
    The proposed change makes the limit on the maximum positive MTC 
more restrictive. From an operational standpoint, a more restrictive 
limit on MTC will help mitigate the effect of plant transients on 
control of plant parameters (e.g., reactor power, pressurizer 
pressure, pressurizer level, etc.) Therefore, the probability of a 
previously analyzed accident will not be significantly increased.
    The reason for the proposed change is to mitigate the effect 
(increased reactor coolant temperatures) of increased SG U-tube 
plugging on the results of the affected safety analyses. Using the 
more restrictive limit on the maximum positive MTC, the Loss of 
Load, Loss of Feedwater Flow, Feed Line Break, and Control Element 
Assembly Withdrawal events were reanalyzed using previously accepted 
methodologies. The results of these analyses are within the 
acceptance limits for these events. Therefore, the consequences of a 
previously analyzed accident will not be significantly increased.
    The proposed change is similar to the examples of amendments 
that are considered not likely to involve significant hazards 
considerations given in the Statements of Consideration for 10 CFR 
50.92 (51 FR 7744). The example of interest is, ``A change that 
constitutes an additional limitation, restriction, or control not 
presently included in the technical specifications, e.g., a more 
stringent surveillance requirement.'' The proposed change provides a 
more restrictive limit on the positive MTC given in Technical 
Specification Figure 3.1.1-1. Based on the above arguments and the 
similarity to an example in the Federal Register, BGE has determined 
that the proposed change does 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.
    The proposed change makes the limit on the maximum positive MTC 
more restrictive. The proposed change does not involve installation 
of new or different equipment, modify the interfaces with existing 
equipment, change the equipment's function, or change the method of 
operating the equipment. The proposed change does not affect normal 
plant operations or configurations. The more restrictive MTC limit 
will help mitigate the effect of plant transients on control of 
plant parameters.
    Therefore, the proposed 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 a margin of 
safety.
    The proposed change provides for a more restrictive limit for 
the allowable positive MTC. The more restrictive limit on the 
maximum positive MTC was evaluated using previously approved 
methodologies and compared to the existing acceptance criteria. The 
analyses show that the proposed change preserves the margin of 
safety by ensuring that the results of the safety analyses for the 
Loss of Load, Loss of Feedwater Flow, Feed Line Break, and Control 
Element Assembly Withdrawal events meet established NRC acceptance 
limits for these events.
    In addition, this proposed change is similar to the example of 
amendments that are considered not likely to involve significant 
hazards considerations given in the Statements of Consideration for 
10 CFR 50.92 (51 FR 7744). The example of interest is, ``A change 
that constitutes an additional limitation, restriction, or control 
not presently included in the technical specifications, e.g., a more 
stringent surveillance requirement.'' The proposed change provides a 
more restrictive limit on the positive MTC given in Technical 
Specification Figure 3.1.1-1. Based on the above arguments and the 
similarity to an example in the Federal Register, BGE has determined 
that the proposed change does not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Susan F. Shankman, Acting.

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

    Date of amendment request: February 8, 1996.
    Description of amendment request: The proposed amendment would 
remove Technical Specifications (TS) 3.3.4, Turbine Overspeed 
Protection; TS 3.7.12, Area Temperature Monitoring; and TS 3.11.2.6, 
Gas Storage Tanks; and their associated bases; and relocate them to 
licensee-controlled documents, such as the Final Safety Analysis 
Report. The licensee revised the original amendment request dated 
October 24, 1994, to provide supplemental information to TS 6.8.4 for 
administrative control program related to TS 3.11.2.6, by letters dated 
August 31, 1995 and February 8, 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 was previously presented in the Federal Register 
(59 FR 60397). The staff reviewed and determined that the proposed 
license amendment's revisions do not alter the original conclusion that 
no significant hazards considerations exist pursuant to 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Eugene V. Imbro.

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

    Date of amendment request: December 21, 1995.
    Description of amendment request: The proposed amendments would 
delete the requirement to place the reactor mode switch in the Shutdown 
position if a stuck open safety/relief valve cannot be closed within 
two minutes. The operator would still be required to scram the reactor 
if suppression pool average water temperature reaches 110 degrees 
Fahrenheit or greater. The licensee also proposed changes to the TS 
index pages to reflect Bases page changes that were accepted by the NRC 
staff in a letter dated May 23, 1995. Because the changes to the index 
pages require a license amendment, they have been included as part of 
this submittal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed change does not involve a significant increase in 
the probability or

[[Page 20845]]

consequences of an accident previously evaluated in the UFSAR. A 
stuck open SRV event is a mild transient which neither affects fuel 
limits nor radiological consequences. The two minute requirement to 
manually scram after a SRV becomes stuck open is not assumed or used 
in any transient or accident analysis in the FSAR. Removing the two 
minute requirement to manually scram after a SRV becomes stuck open 
does not change the probability of any accident evaluated in the 
FSAR. Removing the two minute requirement to manually scram after a 
SRV becomes stuck open also does not change the capability of the 
suppression pool during this event in case of any accident involving 
reactor blowdown, because the suppression pool average water 
temperature limit in Technical Specification 3.6.2.1 is still valid 
and enforced. The suppression pool average water temperature limit 
is the only requirement during operational conditions 1 and 2 that 
assures sufficient heat sink capacity in case of a LOCA in the 
containment. Therefore, removing the two minute requirement to 
manually scram after a SRV becomes stuck open would not increase the 
probability or consequences of any postulated accident analyzed in 
the FSAR.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
in the UFSAR. This change does not effect any hardware. This is a 
procedural change to assure that the reactor will not be 
unnecessarily scrammed by the operator after a SRV is stuck open for 
two minutes. The reactor will still be scrammed if suppression pool 
average water temperature increases above 110 degrees F. Since the 
design basis of the suppression pool is protected by this average 
water temperature limit, this procedural change of removing the two 
minute requirement to manually scram after a SRV becomes stuck open 
introduces no new accident or malfunction.
    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed change does not reduce the margin as defined in the 
bases for any Technical Specification. On the contrary, if the two 
minute requirement to manually scram after a SRV becomes stuck open 
is not removed, the operator has to scram the reactor thus 
challenging the RPS, the rector vessel, and other associated 
components, and reducing the related margin to safety. This scram 
would be unnecessary if the suppression pool average water 
temperature is below the 110 degree F limit allowed by the design 
basis of the suppression pool. Reactor safety or suppression pool 
design basis is not compromised because the suppression pool average 
water temperature limit alone guarantees that there would not be any 
reduction in 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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: March 4, 1996.
    Description of amendment request: The proposed amendments would 
change the McGuire Units 1 and 2 Updated Final Safety Analysis Report 
to delete the seismic qualification requirement for the Containment 
Atmosphere Particulate Radiation Monitors.
    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:


    This proposed change has been evaluated against the standards in 
10 CFR 50.92 and has been determined to involve no significant 
hazards considerations, in that operation of the facility in 
accordance with the proposed amendment would not:
    1. [I]nvolve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    EMF38(L) is not used directly for any phase of power generation 
or conversion or transmission, normal decay heat removal, fuel 
handling, or the processing of radioactive fluids. As such, it is 
not an ``accident initiator''. No ``accident initiator'' is affected 
by the change. Thus, the probability of accidents evaluated in the 
FSAR is not affected by the change. It is determined that sufficient 
ability to determine conditions inside containment remain available 
for any earthquake up to and including the SSE [safe-shutdown 
earthquake]. Furthermore, should either EMF38(L) or EMF39(L) be 
found to not be functional following any earthquake, including those 
smaller than the OBE [Operating Basis Earthquake], the appropriate 
steps will be taken; i.e., declare the monitor(s) inoperable and 
apply the action statement for TS [technical specification] 3.4.6.1 
which may require that the associated unit(s) be taken to Cold 
Shutdown (Mode 5) if the minimum required Reactor Coolant Leakage 
Detection Systems are not operable. Cold Shutdown is a mode for 
which neither the Emergency Core Cooling System nor the containment 
safeguards are required. Finally, no equipment provided to mitigate 
any accident is adversely affected by the change. For these reasons, 
the proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated in 
the SAR [safety analysis report].
    2. [C]reate the possibility of a new or different type of 
accident from any accident previously evaluated; or
    As stated above, no equipment used in direct support of power 
generation or conversion or transmission, normal decay heat removal, 
fuel handling, or processing of radioactive fluids is affected with 
the update. No new failure modes are identified with the change. The 
upper bound to an undetected leak in the Reactor Coolant System is a 
Loss of Coolant Accident [LOCA]. As noted above, no equipment 
provided to mitigate a LOCA is affected by the change. For these 
reasons, the change will not create a new or different type of 
accident from any accident previously evaluated.
    3. [I]nvolve a significant reduction in a margin of safety.
    It has been determined that sufficient means remain at the 
disposal to the operators to assess conditions within the 
containment following any earthquake up to and including the SSE. In 
particular, the ability to determine leakage with the sensitivity 
comparable to that of EMF38(L) can be established. This meets the 
intent of the Regulatory Position of RG [Regulatory Guide] 1.45. In 
addition, should it be determined that either EMF38(L) or EMF39(L) 
is not functional following any earthquake, the appropriate steps 
will be taken; i.e, declare the monitor(s) inoperable and apply the 
action statement for TS 3.4.6.1 which may require that the 
associated unit(s) be taken to Cold Shutdown (Mode 5) if the minimum 
required Reactor Coolant Leakage Detection Systems are not operable. 
This brings the unit(s) to a mode in which TS 3.4.6.1 does not 
apply. It ensures that at least the minimum required Reactor Coolant 
System leakage detection systems will be functional before power 
operations are continued following a postulated earthquake smaller 
than the OBE. It ensures protection of the reactor coolant pressure 
boundary, one of the fission product barriers. No other fission 
product barrier is affected by the change. Therefore, the margin of 
safety is not reduced.
    Therefore, based on the information contained in this submittal, 
it is determined that no significant hazard is associated with the 
proposed change.


    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

[[Page 20846]]

Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas

    Date of amendment request: April 11, 1996.
    Description of amendment request: The proposed technical 
specification amendment modifies the reactor building leak testing 
requirements per Option B to 10 CFR 50, Appendix J. Option B permits 
performance based determination of the reactor building leak testing 
frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    The proposed changes to the Technical Specifications implement 
Option B of 10 CFR 50 Appendix J at ANO. The proposed changes will 
result in increased intervals between containment leakage tests 
determined through a performance based approach. The intervals 
between such tests are not related to conditions which cause 
accidents. The proposed changes do not involve a change to the plant 
design or operation. Therefore, this change does not involve a 
significant increase in the probability of any accident previously 
evaluated.
    NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' 
contributed to the technical bases for Option B of 10 CFR 50 
Appendix J. NUREG-1493 contains a detailed evaluation of the 
expected leakage from containment and the associated consequences. 
The increased risk due to lengthening of the intervals between 
containment leakage tests was also evaluated and found acceptable. 
Using a statistical approach, NUREG-1493 determined the increase in 
the expected dose to the public from extending the testing frequency 
is extremely small. It also concluded that a small increase is 
justifiable due to the benefits which accrue from the interval 
extension. The primary benefit is in the reduction in occupational 
exposure. The reduction in the occupational exposure is a real 
reduction, while the small increase to the public is statistically 
derived using conservative assumptions. Therefore, this change does 
not involve a significant increase in the consequences of any 
accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    The proposed change to the Technical Specifications incorporates 
the performance based approach authorized by Option B of 10 CFR 50 
Appendix J. The interval extensions allowed by this change do not 
involve a change to the plant design or operation. No safety related 
equipment or safety functions are altered as a result of this 
change. The reduced testing frequency does not affect the testing 
methodology. As a result, the proposed change does not affect any of 
the parameters or conditions that could contribute to initiation of 
any accidents. No new accident modes are created by extending the 
test intervals. Therefore, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The proposed change does not change the performance methodology 
of the containment leakage rate testing program. However, the 
proposed change does affect the frequency of containment leakage 
rate testing. With an increased frequency between tests, the 
proposed change does increase the probability that a increase in 
leakage could go undetected for a longer period of time. Operational 
experience has demonstrated the leak tightness of the containment 
buildings has been significantly below the allowable leakage limit.
    The margin to safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rates. The limitation on containment leakage rate is 
designed to ensure the BWN total leakage volume will not exceed the 
value assumed in our accident analysis. The margin to [sic] safety 
for the offsite dose consequences of postulated accidents directly 
related to containment leakage is maintained by meeting the 1.0 L. 
acceptance criteria. The proposed change maintains the 1.0 L. 
acceptance criteria.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: April 11, 1996.
    Description of amendment request: The proposed technical 
specification (TS) amendment adds low-temperature overpressure 
protection (LTOP) requirements to the TSs to resolve Generic Issue 94 
in accordance with Generic Letter 90-06.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    This proposed change provides additional controls in the ANO-2 
Technical Specification [(TS)] for ensuring that LTOP [low-
temperature overpressure protection] protection is available when 
required. The limiting condition involving the simultaneous 
injection of two HPSI [high pressure safety injection] and three 
charging pumps to an RCS [reactor coolant system] water solid 
condition, was used in the calculation of the ANO-2 proposed LTOP 
setpoints. The methodology utilized in the LTOP setpoint analysis is 
based on ASME [American Society of Mechanical Engineers] Code Case 
N-514. The code case establishes a factor of 110 percent of the 
operating pressure temperature curves instead of 100 percent. The 
safety factor utilized by the code case provides a more reasonable 
vessel overpressure allowance for conditions expected under pressure 
loading from low temperature transients. The SITs [safety injection 
tanks] are required to be isolated, if not depressurized, prior to 
entering the LTOP enable temperature and are periodically verified 
to be isolated when LTOP conditions exist. The LTOP setpoint of the 
relief valves proposed by this technical specification [TS] change 
is not considered to be an initiator of any transients, but is used 
to mitigate an overpressure condition if such a transient were to 
occur.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    The design basis event for establishing LTOP limits is the 
simultaneous injection of two HPSI and three charging pumps to an 
RCS water solid condition. The LTOP vent size of 6.38 square inches 
and the valve pressure setpoint of less than or equal to 430 psig 
are currently used for mitigation of low temperature overpressure 
conditions. The change in the enable setpoint was analyzed by the 
application of Code Case N-514 and determined to adequately ensure 
that this temperature [sic] setpoint will mitigate a

[[Page 20847]]

LTOP transient. The operator action to enable the LTOP relief valves 
at 220 degrees ensures that the RCS including the reactor vessel 
will not undergo system pressures at low temperature conditions 
beyond their design limits. Therefore, there will not be any impact 
to systems, structures or components beyond their design 
requirements.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The addition of a new specification to the ANO-2 Technical 
Specification [TS] will not significantly reduce the margin of 
safety. The LTOP safety factors are based on reanalyzed conditions 
for 21 effective full power years (EFPY) of operation utilizing 
methodology contained in ASME Code Case N-514. The LTOP evaluation 
under Code Case N-514 for low temperature transients is considered 
more appropriate than the ASME Section XI. The code case establishes 
a factor of 110 percent of the operating pressure temperature curves 
instead of 100 percent. The safety factor utilized by the code case 
provides a more reasonable vessel overpressure allowance for 
conditions expected under pressure loading from low temperature 
transients. Although the proposed setpoint may involve a slight 
reduction in a margin of safety, the enable temperature setpoint 
will provide an equivalent level of safety to the reactor vessel 
during LTOP transients and will satisfy the purpose of 10 CFR 50.60 
for fracture toughness. Therefore, based on the refined methodology 
used to calculate ANO-2 LTOP setpoints for 21 EFPY the margin of 
safety will not be significantly reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf Nuclear 
Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: April 18, 1996.
    Description of amendment request: The licensee has proposed to 
delete a restriction on the 24-hour emergency diesel generator 
operation test in Surveillance Requirement 3.8.1.14 (Page 3.8-12) of 
the Technical Specifications (TSs) for the Grand Gulf Nuclear Station, 
Unit 1. The deletion would allow the test to also be conducted during 
power operation (i.e., during Modes 1 and 2), instead of the current 
requirement to only conduct the test when the plant is shut down.
    The frequency of conducting this test, the conditions of the test, 
and the criteria to pass the test are not being changed by this 
amendment request.
    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 for the amendment request, which is presented below:

    Entergy Operations, Inc. [(EOI)] propose[d] to change the 
current Grand Gulf Nuclear Station [GGNS] Technical Specifications 
[(TSs)]. The specific change is to modify note 2 to Surveillance 
3.8.1.14. Presently, this note prohibits the performance of the 24 
hour diesel maintenance run while the unit is in either Mode 1 or 2. 
The proposed change would remove this restriction thus allowing the 
24 hour run to be performed during any mode of operation (i.e., 
modes 1, 2, 3, 4 or 5).
    The Commission has provided standards for determining whether a 
no significant hazards considerations exists as stated in 10 CFR 
50.92 (c). A proposed amendment to an operating license involves no 
significant hazards consideration if 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; (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.
    Entergy Operations, Inc. [EOI] has evaluated the no significant 
hazards consideration in its request for this license amendment and 
determined that no significant hazards considerations results from 
this change. In accordance with 10 CFR 50.91(a), Entergy Operations, 
Inc. [EOI] is providing the analysis of the proposed amendment 
against the three standards in 10 CFR 50.92(c). A description of the 
no significant hazards consideration determination follows:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The GGNS UFSAR [Updated Final Safety Analysis Report] assumes 
that the AC electrical power sources are designed to provide 
sufficient capacity, capability, redundancy and reliability to 
ensure that the fuel, reactor coolant system and containment design 
limits are not exceeded during an assumed design basis event. 
Specifically, the UFSAR assumes that the onsite EDG's [emergency 
diesel generator's] provide emergency power in the event offsite 
power is lost to either one or all three ESF [engineered safety 
feature] buses. In the event of a loss of preferred power, the ESF 
electrical loads are automatically connected to the EDG's in 
sufficient time to provide for safe reactor shutdown and to mitigate 
the consequences of a design basis accident such as a LOCA.
    The proposed change to permit the 24 hour testing of the EDG's 
during power operation does not increase the chances or consequences 
of any previously evaluated accident. The capability of the EDG's to 
supply power in a timely manner will not be compromised by 
permitting performance of EDG testing during periods of power 
operation. Design features of the EDG's and electrical systems 
ensures that if a LOCA [loss of coolant accident] or LOP [loss of 
offsite power] signal, either individually or concurrently, should 
occur during testing that the EDG would be returned to its ready-to-
load operation (i.e., EDG running at rated speed and voltage 
separated from the offsite sources) or separately connected to the 
ESF bus providing ESF loads. As such, an EDG being tested is 
considered to be Operable and fully capable of meeting its intended 
design function. Additionally, the testing of an EDG is not a 
precursor to any previously evaluated accidents.
    Therefore, the proposed change allowing testing of EDG's during 
power operation will not significantly increase the probability or 
consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As previously discussed [above], the proposed change to permit 
the performance of EDG testing during power operation will not 
affect the operation of any system or alter any system's response to 
previously evaluated design basis events. The EDG's will 
automatically transfer from the test configuration to the ready-to-
load configuration following receipt of a valid signal (i.e., LOCA 
or LOP). In the ready-to-load configuration, the EDG will be running 
at rated speed and voltage separated from the offsite source capable 
of automatically supplying power to the ESF buses in the event that 
preferred power is actually loss.
    Surveillance Requirement 3.8.1.17 demonstrates that the EDG will 
automatically override the test mode following generation of a LOCA 
signal. In addition the ability of the EDG's to survive a full load 
reject is verified by the performance of surveillance requirement 
3.8.1.10. These existing surveillance requirements along with system 
design features ensures that the performance of EDG testing during 
power operation will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    The AC electrical power sources are designed to provide 
sufficient capacity, capability, redundancy, and reliability to 
ensure the availability of necessary power to ESF systems so that 
the fuel, reactor coolant system and containment design limits are 
not exceeded. Specifically, the EDG's must be capable of 
automatically providing power to ESF loads in sufficient time to 
provide for safe reactor shutdown and to mitigate the consequences 
of a design basis accident in the event of a loss of preferred 
power.

[[Page 20848]]

    Testing of EDG's during power operation will not affect the 
availability or operation of any offsite source of power. In 
addition, the EDG being tested remains capable of meeting its 
intended design functions. Therefore the proposed change to the 
Technical Specification Surveillance Requirement 3.8.1.14 will not 
result in a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: April 15, 1996 (TSCR No. 244).
    Description of amendment request: The proposed amendment would 
revise Specification 5.3.1.B of the Oyster Creek Technical 
Specifications. The current specification prohibits handling a load 
greater in weight than one fuel assembly over irradiated fuel in the 
spent fuel storage facility. The proposed change will facilitate the 
off load of spent fuel to the Oyster Creek Independent Spent Fuel 
Storage Installation (ISFSI). Specifically, the shield plug for the dry 
shield canister (DSC) and the associated lifting hardware will be moved 
over irradiated fuel which is contained in the DSC within the transfer 
cask located in the Cask Drop Protection System (CDPS).
    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. State the basis for the determination that the proposed 
activity will or will not increase the probability of occurrence or 
consequences of an accident.
    The design features and capacity of the reactor building crane 
provide a significant safety factor. In addition, personnel training 
and other administrative controls further reduce risk. Thus, the 
dropping of the DSC shield plug onto a loaded DSC and causing damage 
to the spent fuel assemblies is not a credible event. Therefore, it 
does not increase the probability of or consequences of an accident.
    2. State the basis for the determination that the activity does 
or does not create the possibility of an accident or malfunction of 
a different type than any previously identified in the SAR [safety 
analysis report].
    This activity will not create the possibility of a new or 
different type of accident than previously evaluated in the SAR 
because the proposed heavy load handling exception does not create a 
new credible accident scenario. Dropping the shield plug on a loaded 
DSC and damaging spent fuel assemblies is not considered a credible 
event.
    3. State the basis for the determination that the margin of 
safety is not reduced.
    This activity will not involve a significant reduction in the 
margin of safety because the proposed heavy load handling evolution 
does not create a credible accident scenario.

    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.

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

    Date of amendment request: April 19, 1996.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.14 to add the appropriate references 
identifying the detailed methodology and conditions for analyzing the 
Small Break Loss-of-Coolant Accident (SBLOCA) to the list of the 
approved Core Operating Limits Report methods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the Proposed Amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    These Proposed Changes are administrative in nature and are 
consistent with the guidance set forth in the NRC Generic Letter 88-
16 identifying the requirements for the inclusion of analytical 
methodology references in Technical Specifications as used in 
determining compliance with the regulatory limits.
    The references, as proposed to be included in section 5.14 of 
the Technical Specifications, have previously been reviewed and 
approved by the NRC for generic applicability to PWRs [Pressurized 
Water Reactors]. The reports identified in the Proposed Change have 
been accepted by the NRC for referencing in plant licensing 
applications.
    Since the references listed in the Proposed Change have 
previously been found to meet the conditions of 10 CFR 50.46 and 10 
CFR Appendix K, and that the plant specific safety analysis 
acceptance limits have not changed or been modified, the use of 
these references in the analysis of SBLOCA accident for the Maine 
Yankee plant is consistent with prior plant specific and industry 
requirements and practices.
    Therefore, we have concluded that the Proposed Change will not 
result in a significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Does the Proposed Amendment create the possibility for a new 
or different kind of accident?
    The Proposed Changes introduce no new mode of plant operation; 
do not involve the physical modification of any structure, system, 
or component; do not affect the function, operation or surveillance 
for any equipment necessary for safe operation or shutdown of the 
plant; and, do not involve any changes to setpoints or limits or 
operating parameters. The Proposed Changes are administrative in 
nature only.
    Therefore, we have concluded that the Proposed Change cannot 
result in the possibility of a new or different kind of accident 
from that previously evaluated.
    3. Does the Proposed Amendment involve a significant reduction 
in a margin of safety?
    The Proposed Changes are administrative in nature, consistent 
with the guidance of Generic Letter 88-12, and have been reviewed 
previously by the NRC and found acceptable with regard to the 
requirements of 10 CFR 50.46 and 10 CFR Appendix K. Additionally, 
the plant specific safety analysis acceptance criteria has not 
changed from that used in the latest core reload analysis.
    Therefore, we have concluded that the Proposed Change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011.

[[Page 20849]]

    NRC Deputy Director: John A. Zwolinski.

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

    Date of amendment request: February 7, 1996.
    Description of amendment request: The amendment would change the 
operating license, the Technical Specifications, and associated Bases 
to permit the use of 10 CFR Part 50, Appendix J, Option B, Performance-
Based Containment Leakage Rate Testing in accordance with the 
implementation guidance in NRC's Regulatory Guide 1.163 dated September 
1995. The change to the operating license would delete, in paragraph 
2.D.ii, reference to certain exemptions to Appendix J previously 
granted by the NRC, which would no longer be applicable once Option B 
is implemented.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    NMP2 [Nine Mile Point, Unit 2] is currently implementing Option 
A of Appendix J of 10 CFR 50 for Type A, B and C testing. The 
proposed change to the Operating License, the Technical 
Specifications and the Bases would implement Option B to Appendix J 
of 10 CFR 50 at NMP2 for Type A, B and C testing. Option B would 
allow increased testing intervals after satisfying certain 
performance based criteria. The proposed change also corrects an 
inconsistency between the restoration statements and the 
applicability requirements of LCO [Limiting Condition of Operation] 
3.6.1.2. In addition, the proposed change affects the testing 
intervals for the verification of the interlocks on the primary 
containment air lock and for the measuring of the Hydrogen 
Recombiner System leakage rate.
    Appendix J describes the requirements for leakage testing of the 
primary containment and its components penetrating the primary 
containment. The leakage testing interval of the primary containment 
and its components is not a precursor or initiator to an accident. 
The primary containment and its penetrations minimizes the leakage 
of radioactivity into the environment during an accident which 
pressurizes the primary containment.
    The testing intervals of the air lock interlocks and of the 
Hydrogen Recombiner System leakage rate are also not precursors or 
initiators to an accident. The interlocks function to provide 
assurance that at least one air lock door will be closed and thereby 
perform its accident mitigating function of minimizing the leakage 
of radioactivity into the environment during accident conditions. 
The Hydrogen Recombiner System is manually initiated following a 
loss-of-coolant accident (LOCA) to maintain the hydrogen 
concentration within the primary containment below its flammable 
limit during post-LOCA conditions.
    An inconsistency exists between the applicability statement of 
LCO 3.6.1.2 and the requirement of the restoration statements to 
restore prior to increasing reactor coolant system temperature over 
200  deg.F. Eliminating this inconsistency does not diminish the 
requirements contained in the Technical Specifications.
    Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    The proposed change to the Operating License, the Technical 
Specifications and the Bases would replace the detailed and 
prescriptive technical requirements contained in Option A of 
Appendix J with performance based requirements and supporting 
regulatory/industry documents contained in Option B of Appendix J. 
This proposed change includes a description of the 10 CFR 50 
Appendix J Testing Program Plan in Section 6.8.4.f of the Technical 
Specifications.
    This program plan, with one exception, is consistent with RG 
[Regulatory Guide] 1.163. This exception to the RG is acceptable as 
it is technically equivalent to and replaces an exemption that was 
applicable to Option A of Appendix J. Therefore, this program plan 
establishes leakage-rate test methods, procedures, acceptance 
criteria and analyses which comply with Option B of Appendix J to 10 
CFR 50.
    The implementation of this program continues to provide adequate 
assurance that during a DBA [Design Basis Accident]-LOCA the primary 
containment and its components will continue to limit leakage rates 
to less than the allowable leakage rates described in the Technical 
Specifications and thereby limit leakage consistent with the 
assumptions of the accident analyses. Therefore, the increased test 
intervals permitted by Option B for the primary containment and its 
penetrations will continue to implement the safety objectives 
underlying the requirements of Appendix J.
    As discussed under the margin of safety, the impact of the 
proposed change on the consequences of a release is negligible. The 
slight increase in the risk to the population is compensated by the 
corresponding risk reduction benefits associated with the reduction 
in component cycling, stress, and wear associated with increased 
test intervals.
    At least one air lock door in each air lock will continue to be 
closed during the onset of an accident that would release 
radioactivity into primary containment. Therefore, the air lock 
interlocks continue to provide assurance that at least one leak 
tested barrier will limit leakage during accident conditions.
    The Hydrogen Recombiner System will continue to operate to 
maintain the hydrogen concentration within the primary containment 
below its flammable limit during post-LOCA conditions. This provides 
assurance that primary containment integrity will not be challenged 
by hydrogen burns.
    Eliminating the inconsistency between the restoration statements 
and the applicability requirements of LCO 3.6.1.2 does not diminish 
the requirements contained in the Technical Specifications. The 
Technical Specifications continue to require that the leakage limits 
of LCO 3.6.1.2 be met prior to entering OPERATIONAL CONDITIONS 1, 2, 
or 3 (i.e., temperature greater than 200  deg.F).
    Accordingly, operation with the proposed change to the Operating 
License, the Technical Specifications and the Bases will not 
significantly increase the consequences of an accident previously 
evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change would implement Option B of Appendix J of 10 
CFR 50 for Type A, B and C testing. Option B would allow increased 
testing intervals after satisfying certain performance based 
criteria. The proposed change also corrects any inconsistency 
between the restoration statements and the applicability 
requirements of LCO 3.6.1.2. In addition, the proposed change 
affects the testing intervals for the interlocks on the primary 
containment air lock and for the measuring of the Hydrogen 
Recombiner System leakage rate.
    No new plant operating modes, system operating configurations 
nor failure modes are introduced by the proposed change. The primary 
containment and its penetrations will continue to perform their 
accident mitigating function. The Hydrogen Recombiner System will 
continue to function to prevent hydrogen burns within primary 
containment during post-LOCA conditions.
    Accordingly, operation with the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    A regulatory impact analysis of implementing performance-based 
requirements indicates that relaxing the frequency of Type A, B and 
C testing leads to an increase in overall reactor risk of 
approximately two percent. As indicated in the Staff's Regulatory 
Impact Analysis, this increase is considered to be marginal to 
safety.
    As indicated above, increasing test intervals can slightly 
increase the risk to the population associated with the consequences 
of a release; however, this is compensated by the corresponding risk 
reduction benefits associated with the reduction in component 
cycling, stress, and wear associated with increased test intervals. 
Therefore, when considering the total integrated risk, the risk 
associated with increased test intervals is negligible.

[[Page 20850]]

    The proposed change is consistent with current plant safety 
analyses. In addition, the proposed change does not require 
revisions to the design of NMP2. As such, the proposed individual 
changes will maintain the same level of reliability of the equipment 
associated with containment integrity, assumed to operate in the 
plant safety analysis, or provide continued assurance that specified 
parameters affecting leak rate integrity, will remain within their 
acceptance limits.
    The as-left leakage after performing a required leakage test 
continues to be less than 0.60 La for combined Type B and C leakage 
and less than or equal to 0.75 La for Type A leakage. These as-left 
acceptance criteria and the testing frequency as established by the 
10 CFR 50 Appendix J Testing Program Plan provide assurance that the 
measured leakage rate will not exceed the maximum allowable leakage 
of La during plant operation.
    Visual examination of accessible interior and exterior surfaces 
of the primary containment continues to be performed prior to 
initiating a Type A test. The total number of visual examinations 
performed will continue to be three times during a 10-year period. 
Therefore, visual examinations of the primary containment will 
continue to allow for the timely uncovering of evidence of 
structural deterioration and satisfy the requirements of RG 1.163.
    The primary containment air lock interlocks will be tested prior 
to conducting an air lock seal leakage test. This testing 
requirement continues to provide adequate assurance that at least 
one leak tested air lock door in each air lock will be closed during 
accident conditions.
    The measuring of the Hydrogen Recombiner System Leakage rate 
will continue to be included as part of the overall integrated 
leakage rate test. The test schedule for measuring system leakage 
will also continue to coincide with the schedule for performing a 
Type A test.
    The leakage limits of LCO 3.6.1.2 will continue to be met prior 
to entering into OPERATIONAL CONDITIONS 1, 2, or 3 (i.e., 
temperature greater than 200  deg.F). Satisfying these leakage 
limits provides assurance that the measured leakage rate will not 
exceed the maximum allowable leakage rate of La during plant 
operation. Therefore, operation with the proposed change will not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: January 17, 1996.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) including revisions to 
Specifications 3/4.3.1, ``Reactor Protection System Instrumentation,'' 
3/4.3.2, ``Isolation Actuation Instrumentation,'' 3/4.3.3, ``Emergency 
Core Cooling System Actuation Instrumentation,'' 3/4.3.4.2, ``End-of-
Cycle Recirculation Pump Trip System Instrumentation,'' and the 
associated Bases to relocate response time limit tables from the TSs to 
the Updated Safety Analysis Report (USAR). The proposed revisions to 
the TSs also include several administrative changes.
    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 operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment relocates Tables 3.3.1-2, ``Reactor 
Protection System Response Times,'' 3.3.2-3, ``Isolation System 
Instrumentation Response Times'' 3.3.3-3, ``Emergency Core Cooling 
System Response Times'' and 3.3.4.2-3 ``End-of-Cycle Recirculation 
Pump Trip System Response Time'' from the Technical Specifications 
to the USAR. The Technical Specification Surveillance Requirements 
and associated actions are not affected and remain in the Technical 
Specifications. This change to the reactor protection system 
instrumentation, isolation actuation instrumentation, and emergency 
core cooling system instrumentation is being done in accordance with 
the guidance provided in Generic Letter 93-08, ``Relocation of 
Technical Specification Tables of Instrument Response Time Limits,'' 
and the change to the end-of-cycle recirculation pump trip system 
instrumentation is consistent with NUREG 1433, ``Standard Technical 
Specifications, BWR/4.'' This change allows NMP2 [Nine Mile Point 
Unit 2] to administratively control subsequent changes to the 
response time limits in accordance with 10CFR50.59. Additionally, 
procedures which contain the various response time limits are also 
subject to the change control provisions of 10 CFR 50.59. Relocating 
this information does not affect the initial conditions of a design 
basis accident or transient analysis. The proposed Technical 
Specification changes do not affect the capability of the associated 
systems to perform their intended functions within their required 
response times. Since any subsequent changes to the USAR or 
procedures which contain the response time limits are evaluated in 
accordance with 10CFR50.59, the proposed amendment does not involve 
an increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed change would relocate the response time limit 
tables from the Technical Specifications to the USAR. Subsequent 
changes to the USAR, or in procedures which contain the various 
response time limits, would be evaluated in accordance with the 
requirements of 10CFR50.59, which would evaluate the possibility of 
the creation of a new or different kind of accident. The proposed 
change does not involve any physical alteration of the plant, change 
in a Limiting Condition for Operation or change in Surveillance 
Requirements. No new failure modes are introduced. Therefore, this 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The proposed change would relocate the response time limit 
tables from the Technical Specifications to the USAR. Future changes 
to the response time limits in the USAR, or in procedures which 
contain the various response time limits, would be in accordance 
with 10CFR50.59, which would evaluate the proposed change to 
determine whether it involved any reduction in the margin of safety. 
The response time limits to be transposed from the Technical 
Specifications to the USAR are the same as the existing Technical 
Specifications. Therefore, this 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 request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Susan Frant Shankman, Acting.

[[Page 20851]]

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

    Date of amendment request: January 25, 1996.
    Description of amendment request: The proposed amendment would 
change a footnote in Table 3.3.3-1 and the corresponding footnote in 
surveillance Table 4.3.3.1-1 (both referenced by Technical 
Specification 3/4.3.3 ``Emergency Core Cooling System Actuation 
Instrumentation'') to more clearly define when, during cold shutdown 
and refueling (i.e., Operational Conditions 4 and 5), the Loss of 
Voltage and Degraded Voltage relays associated with the 4.16 kV 
Emergency Bus Undervoltage are required to be operable. The footnotes 
currently state: ``Required when ESF [Engineered Safety Features] 
equipment is required to be OPERABLE.'' The proposed amendment would 
change the footnotes to state: ``Required when the associated diesel 
generator is required to be OPERABLE.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    The proposed change would require the Loss of Power instruments 
to be OPERABLE in Operational Conditions 4 and 5 only when the 
associated diesel generator is required to be OPERABLE. The Loss of 
Power relays provide a support function to initiate the associated 
diesel generator start and bus unloading sequences. If that diesel 
generator is not in service, the loss of power relays perform no 
safety function. Therefore, relating diesel generator OPERABILITY 
and Loss of Power instrument OPERABILITY will not involve an 
increase in the probability of an accident previously evaluated.
    The proposed change does not affect the requirements of ESF 
OPERABILITY. The change does not affect diesel generator response to 
a loss of voltage or degraded voltage on the Divisional 4.16 kV 
electrical busses when the diesel generator is required to be 
OPERABLE. Automatic response of the ESF functions is unaffected by 
removing the Loss of Power relays from service under these 
conditions, therefore, the proposed change will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not involve a modification of plant 
equipment nor does it change the way the equipment will be 
maintained or operated. The revision to Technical Specifications 
will continue to require the Loss of Power instrumentation to be 
OPERABLE when the associated diesel generator is required to be 
OPERABLE. The Loss of Power instruments will continue to perform 
their safety function of initiating the diesel generator start and 
bus unloading sequences.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed change will not affect the OPERABILITY, operation 
or reliability of any ESF function including the diesel generators. 
All ESF functions will remain available during postulated accidents 
with a loss of offsite electrical power. The change simply clarifies 
when the Loss of Power instruments are required to be OPERABLE 
during Operational Conditions 4 and 5. Therefore, the proposed 
change will not involve a significant reduction in a margin of 
safety.

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

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

    Date of amendment request: March 15, 1996.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements of Technical Specification (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 test from 18 months to an 
interval corresponding to that required for the Containment Integrated 
Leak Rate Test. The provisions of TS 4.0.2 (which would provide an 
extension of up to 25% of the specified surveillance interval) will not 
apply. Specifically, existing TS 4.6.2.1.d would become subparagraphs d 
and e to require that the suppression pool be demonstrated operable:

    d. At least once per 18 months by conducting a visual inspection 
of the exposed accessible interior and exterior surfaces of the 
suppression chamber.*
    e. At least every outage by requiring the performance of a 
Containment Integrated Leak Rate Test, as scheduled in conformance 
with the criteria specified in the 10 CFR 50 Appendix J Testing 
Program Plan described in Section 6.8.4.f, by conducting a drywell-
to-suppression chamber bypass leak test at an initial differential 
pressure of 3 psi and verifying that the [drywell-to-suppression 
chamber bypass flow area] A/the square root of K calculated from the 
measured leakage is within the specified limit of 0.0054 square 
feet.
    1. If any drywell-to-suppression chamber bypass leak test fails 
to meet the specified limit, the test schedule for subsequent tests 
shall be reviewed and approved by the Commission.
    2. If two consecutive tests fail to meet the specified limit, a 
test shall be performed at least each refueling outage until two 
consecutive tests meet the specified limit, at which time the 
original test schedule may be resumed.
    3. The provisions of Specification 4.0.2 do not apply.

    *Includes each vacuum relief valve and associated piping.
    The proposed changes would also add a new surveillance requirement 
for the testing of the bypass leakage path containing the suppression 
chamber vacuum breakers, with associated acceptance criteria, which 
would be performed each refueling outage that the bypass leak test is 
not performed. Specifically, a new TS 4.6.2.1f would require that the 
suppression pool be demonstrated operable:

    f. During each refueling outage for which the drywell-to-
suppression chamber bypass leak test in Specification 4.6.2.1.e is 
not conducted, by conducting a test of the four drywell-to-
suppression chamber bypass leak paths containing the suppression 
chamber vacuum breakers at a differential pressure of at least 3 psi 
and
    1. Verifying that the total leakage area A/the square root of K 
contributed by all four bypass leak paths is less than or equal to 
24% of the specified limit, and
    2. The leakage area for any one of the four bypass leak paths is 
less than or equal to 12% of the specified limit.

    By separate action, the NRC has provided notice of a proposed 
amendment to change the frequency of Containment Integrated Leak Rate 
Tests in accordance with Option B of 10 CFR Part 50 Appendix J . The 
proposed changes described herein are intended

[[Page 20852]]

to be consistent with the changes proposed under Option B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation on Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed TS changes involve the drywell-to-suppression 
chamber bypass leak test frequency. There are no physical or 
operational changes to the plant as a result of these proposed TS 
revisions. Furthermore, the primary containment acts as an accident 
mitigator and not as an accident initiator. Therefore, the proposed 
TS changes do not affect the probability of any previously evaluated 
accident.
    The continued testing of bypass leakage pathways containing the 
suppression chamber vacuum breakers on a refueling frequency, and 
the continued requirement for visual inspection of containment 
structural features assures that the bypass leakage path will not 
degrade beyond the TS allowable limit during the interval between 
performance of the bypass leakage test. Therefore, radioactivity 
release following an accident will not be increased since the 
pressure suppression capability of the containment is not reduced 
from the existing design, and there will be no significant increase 
in the consequences of any accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed TS changes involve the drywell to suppression 
chamber bypass leak test frequency. There are no physical or 
operational changes as a result of these proposed TS changes. These 
proposed TS changes also include a requirement to continue 
performing a surveillance test on the bypass leakage pathways 
containing the vacuum breaker assemblies each refueling outage for 
which the drywell-to-suppression chamber test is not conducted. This 
test, along with the visual inspection required every refueling 
cycle, will ensure that acceptable bypass leakage is maintained 
during those intervals when the bypass leak test is not required. 
Accordingly, the possibility of a new or different type of accident 
is not introduced. Therefore, the proposed TS changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The drywell-to-suppression chamber bypass leak test data 
obtained during previous testing at NMP2 [Nine Mile Point Unit 2] 
demonstrates conformance, by a large margin, to the TS and design 
leakage requirements. The test data and engineering evaluations 
indicate that there is negligible risk that the bypass leakage will 
change adversely in future years. Furthermore, the proposed test 
frequency is judged to be acceptable based on the small risk of 
bypass leakage through paths other than those containing the 
suppression chamber vacuum breakers.
    A test of the bypass leak pathways containing the vacuum 
breakers will be used to verify acceptable bypass leakage during 
those outages when the bypass leak test is not performed. The 
proposed test of the bypass leak pathways containing the vacuum 
breakers, with stringent acceptance criteria, combined with the 
other negligible potential leakage areas provide an acceptable level 
of assurance that the bypass leakage can be measured. This 
capability ensures that an adverse condition can be detected and 
corrected such that the existing level of confidence that the 
primary containment will function as required during a LOCA [loss-
of-coolant accident] is maintained. Therefore, the proposed TS 
changes do not involve a significant reduction in the margin of 
safety.

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

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

    Date of amendment request: March 20, 1996.
    Description of amendment request: The proposed amendment would 
revise Tables 3.3.1-1 and 4.3.1-1 of Technical Specification 3/4.3.1 
``Reactor Protection System Instrumentation'' to delete the operability 
requirement for the Average Power Range Monitor (APRM) Neutron Flux-
Upscale, Setdown and Inoperative functions in Operational Conditions 
(OCs) 3 (Hot Shutdown) and 4 (Cold Shutdown). These same functions 
would also be revised for OC 5 (Refueling) to indicate that operability 
will only be required during shutdown margin demonstrations performed 
per TS 3.10.3.
    Basis for proposed no significant hazards consideration 
determination: The revisions to the APRM functions are proposed to 
support licensee's plans to replace Local Power Range Monitors during 
the next refueling outage. The revisions also provide for the eventual 
replacement of the existing APRM System with the Nuclear Measurement 
Analysis and Control Power Range Neutron Monitoring System, and the 
eventual installation of the Oscillation Power Range Monitor system for 
the detection of reactor instability conditions. These modifications 
are based upon Report NEDO-31960, ``BWR Owners' Group Long-Term 
Solutions Licensing Methodology, approved by the Commission July 12, 
1993; the licensee's response of November 8, 1994, selecting Option III 
in NEDO-31960 for Nine Mile Point, Unit 2; NRC Generic Letter 94-02, 
``Long-Term Solutions and Upgrade of Interim Operating Recommendations 
for Thermal-Hydraulic Instabilities in Boiling Water Reactors'' dated 
July 11, 1994; and General Electric Licensing Topical Report, NEDC-
32410P-A, ``Nuclear Measurement Analysis and Control Power Range 
Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip 
Function,'' which was approved by the Commission September 5, 1995.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    The operation of Nine Mile Point Unit 2 in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The Reactor Protection System (RPS) initiates a reactor scram 
when one or more monitored parameters exceed their specified limits 
to preserve the integrity of the fuel cladding and the Reactor 
Coolant System and to minimize the energy that must be absorbed 
following a loss-of-coolant accident. The proposed changes will 
revise the OCs in which the APRM Neutron Flux-Upscale, Setdown and 
Inoperative RPS Instrumentation is required. These changes do not 
affect the probability of precursors of any accidents previously 
evaluated, and therefore, do not increase their probability.
    During normal operation in OCs 3 and 4, all control rods are 
fully inserted and the reactor mode switch position control rod 
withdrawal blocks do not allow control rods to be withdrawn. 
Therefore, the RPS APRM functions are not required. Specification 
3.9.10 does allow one control rod to be removed from the core in OC 
4 by placing the mode switch in the refuel position. However, with 
the reactor mode switch in the refuel position, refueling interlocks 
are in place (i.e., one-rod out, etc.), which together with

[[Page 20853]]

adequate shutdown margin will preclude unacceptable reactivity 
excursions. The APRM Neutron Flux-Upscale, Setdown function is not 
required during OC 5 except during shutdown margin demonstrations. 
The SRMs [source range monitors], IRMs [intermediate range 
monitors], and refueling interlocks provide adequate protection from 
reactivity excursions during OC 5. The exception is during the 
shutdown margin demonstration when more than one control rod will be 
withdrawn and the APRMs will continue to be required to be operable 
as a backup to the IRMs. Testing of the RPS APRM functions will 
continue to be performed in those OCs for which operability is 
required. Consequently, the reliability and performance of the RPS 
APRM functions in these OCs will not be adversely affected. 
Therefore, the proposed change will not result in a significant 
increase in the consequences of any accidents previously evaluated.
    The operation of Nine Mile Point Unit 2 in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes will revise the applicable OCs in which the 
APRM neutron Flux-Upscale, Setdown and Inoperative RPS 
instrumentation is required. Changes to OC requirements will not 
introduce any new accident precursors and will not involve any 
physical alternations to plant configurations which could initiate a 
new or different kind of accident. NMP2 is analyzed for a single 
control rod withdrawal error during refueling. Since the core is 
designed to meet shutdown requirements with the highest worth rod 
withdrawn, the core remains subcritical even with one rod withdrawn. 
The one-rod-out interlock which allows only one control rod to be 
withdrawn in OC 5 is not affected by the proposed changes. 
Consequently, the proposed changes do not create an accident 
different than the previously analyzed single control rod withdrawal 
error event. Surveillance testing will continue to be performed to 
assure reliability and maintain current performance levels. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The operation of Nine Mile Point Unit 2 in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.
    The proposed changes to the RPS APRM function instrumentation 
Technical Specification requirements will not adversely affect the 
design or the performance characteristics of the RPS instrumentation 
nor will it affect the ability of the RPS APRM instrumentation to 
perform its intended function. As discussed above, the subject RPS 
instrumentation is not required in OC 3, 4, and 5 except for 
shutdown margin demonstrations. Accordingly, deletion of the 
requirement to have these functions operable in these OCs will not 
significantly reduce a margin of safety. Surveillance testing will 
continue to be performed for those OCs in which the instrumentation 
is required to assure reliability. Therefore, the proposed changes 
do not involve a significant reduction in a margin of safety

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

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

    Date of amendment request: March 28, 1996.
    Description of amendment request: The proposed amendment would 
change Technical Specification Section 3.7.7, ``Sealed Source 
Contamination,'' by making the criteria for testing sealed sources for 
contamination and leakage at Millstone Unit No. 2 the same as those at 
Millstone Unit No. 3, the Haddam Neck Plant, and Seabrook Station. 
Specifically, the sealed sources that are required to be free of 
greater than or equal to 0.005 microcuries of removable contamination 
would be those that would exceed ``100 microcuries of beta and/or gamma 
emitting material or 5 microcuries of alpha emitting material.'' The 
Bases Section 3/4.7.7, ``Sealed Source Contamination,'' would also be 
changed to reference the appropriate section of 10 CFR 70.39.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    Pursuant to 10 CFR 50.92, NNECO [Northeast Nuclear Energy 
Company] has reviewed the proposed changes and concludes that the 
changes do not involve a significant hazards consideration (SHC) 
since the proposed changes satisfy the criteria in 10 CFR 50.92(c). 
That is, the proposed changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The changes make the criteria for testing sealed sources for 
contamination and leakage at Millstone Unit No. 2 the same as those 
at Millstone Unit No. 3, the Haddam Neck Plant and Seabrook Station. 
Although the leakage criteria for sealed sources that are to be 
tested is being changed, the allowable leakage remains small. Any 
leakage that is identified would not cause a significant radiation 
exposure. The source storage area is routinely surveyed by Health 
Physics in accordance with Health Physics Department procedures and 
any significant leakage would be detected. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change in the criteria for testing sealed sources 
for contamination and leakage will not change the way the sources 
are used. Therefore, this change will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The possible radiation exposure to both the workers and the 
public from this change is very small. All protective systems which 
would detect any release of material from the site remain in place 
so there is no reduction in safety for the public. Likewise, all 
protective systems for the workers remain in place. Workers using 
the sources routinely pass through the whole body contamination 
monitors. In addition, the source storage areas are surveyed 
routinely by Health Physics in accordance with Health Physics 
Department procedures, and any significant leakage would be 
detected. The bases section is being revised to reference the 
appropriate section of 10 CFR 70.39. Therefore, there is no 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and Waterford Library, Attn: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT 06385.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: March 12, 1996.

[[Page 20854]]

    Description of amendment request: The proposed changes would remove 
a requirement to interconnect two or more accumulators for the purpose 
of cross checking instrumentation in the event that one of the two 
pressure or level instrument channels on an accumulator is declared 
inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: The design basis accident for which the accumulators 
were designed is the double ended guillotine break of a cold leg. 
Interconnecting or not interconnecting accumulators does not have 
any effect on the probability of occurrence of this event. By 
eliminating the requirement to interconnect accumulators, the 
proposed amendment assures that a minimum of three accumulators are 
available, as assumed in the safety analyses, to mitigate the 
consequences of a large-break loss-of-coolant [LBLOCA] accident. 
Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed amendment does not involve any physical 
changes to plant equipment or setpoints and does not create the 
possibility of a new or different kind of accident. Eliminating the 
requirement to interconnect accumulators ensures that the plant 
configuration is maintained consistent with that assumed in the 
safety analysis and no new failure modes are created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?

    Response: There is no margin of safety specified in the 
Technical Specifications for these instrument channels. There are no 
setpoints or allowable values associated with these instrument 
channels which affect Safety Limits or Limiting Safety System 
Settings. The proposed amendment ensures that the safety analysis 
assumption regarding the accumulators remains valid and the 
resulting peak fuel clad temperature meets specified acceptance 
criteria. The proposed amendment 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 request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Susan Frant Shankman, Acting.

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

    Date of amendment request: March 14, 1996.
    Description of amendment request: The proposed changes would allow 
a one-time extension of the inspection interval for the steam generator 
tubes that is due in July 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) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: The proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. As stated in the Basis of the IP3 
[Indian Point Unit 3] Technical Specifications, the program for 
inservice inspection of steam generator tubes regarding equipment, 
procedures, and sample selection is based upon the guidance and 
recommendations in Regulatory Guide 1.83 and NRC Generic Letter 85-
02. The addition of the footnote to extend the surveillance due date 
will not increase the deviation from the guidance and recommendation 
stated above, and, therefore will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed change does not involve the 
addition of any new or different type of equipment, nor does it 
involve the operation of equipment required for safe operation of 
the facility in a manner different from those addressed in the Final 
Safety Analysis Report. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: The proposed license amendment does not involve a 
significant reduction in a margin of safety. The proposed change 
does not adversely affect any safety related system or component 
operation or operability, instrument operation, or safety system 
setpoints and does not result in increased severity of any of the 
accidents considered in the safety analysis. This change has no 
adverse effect on any margin of safety and, therefore, does not 
create a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Susan Frant Shankman, Acting

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

    Date of amendment request: March 22, 1996.
    Description of amendment request: The amendment proposes changes to 
the Technical Specifications (TS) to establish operability requirements 
for avoidance and protection from thermal hydraulic instabilities to be 
consistent with Boiling Water Reactor Owners Group long-term solution 
Option I-D. Editorial changes are also made to support the revised 
specifications, improve readability of Bases sections, and enhance the 
presentation of requirements for single loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The implementation of BWR Owners' Group long-term stability 
solution Option I-D at FitzPatrick does not modify the assumptions 
contained in the existing accident analysis. The use of an exclusion 
region and the operator actions required to avoid and minimize 
operation inside the region do not increase the possibility of an 
accident. Conditions of operation outside of the exclusion region 
are within the analytical envelope of the existing safety analysis. 
The operator action requirement to exit the

[[Page 20855]]

exclusion region upon entry minimizes the possibility of an 
oscillation occurring. The actions to drive control rods and/or to 
increase recirculation flow to exit the region are maneuvers within 
the envelope of normal plant evolutions. The flow referenced scram 
has been analyzed and will provide automatic fuel protection in the 
event of an instability. Thus, each proposed operating requirement 
provides defense in depth for protection from an instability event 
while maintaining the existing assumptions of the accident analysis.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated because:
    The proposed operating requirements either mandate operation 
within the envelope of existing plant operating conditions or force 
specific operating maneuvers within those carried out in normal 
operation. Since operation of the plant with all of the proposed 
requirements are within the existing operating basis, an unanalyzed 
accident will not be created through implementation of the proposed 
change.
    3. Involve a significant reduction in the margin of safety 
because:
    Each of the proposed requirements for plant thermal hydraulic 
stability provides a means for fuel protection. The combination of 
avoiding possible unstable conditions and the automatic flow 
referenced reactor scram provides an in depth means for fuel 
protection. Therefore, the individual or combination of means to 
avoid and suppress an instability supplements the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Susan Frank Shankman, Acting.

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

    Date of amendment request: March 22, 1996.
    Description of amendment request: The amendment proposes to revise 
Technical Specification (TS) Table 3.2-2, ``Core and Containment 
Cooling System Initiation and Control Instrumentation Operability 
Requirements.'' The proposed changes will revise allowed outage times 
(AOTs) for 4kV Emergency Bus Undervoltage Trip Functions. The AOTs for 
these trip functions were extended by Amendment 227; however, the AOT 
extensions for these trip functions were not consistent with the 
requirements of Standard Technical Specifications (STS), NUREG-1433, 
and differed from the recommendations in the associated Licensing 
Topical Report. Additional changes are proposed to TS Table 3.2-2 and 
to TS Table 4.2-2, ``Core and Containment Cooling System 
Instrumentation Test and Calibration Requirements.'' These changes 
will: (1) replace the generic actions for inoperable instrument 
channels with function-specific actions, (2) replace the generic test 
AOT with function-specific test AOTs, and (3) relocate selected trip 
functions from the TS to an Authority controlled document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are limited to replacement of the generic 
actions and test AOT with function-specific actions and test AOTs, 
and relocation of selected trip functions from the TS to an 
Authority controlled document. The changes do not introduce any new 
modes of plant operation, make any physical changes, or alter any 
operational setpoints. Therefore, the changes do not degrade the 
performance of any safety system assumed to function in the accident 
analysis. Consequently, there is no effect on the probability or 
consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms since the changes do not introduce 
any new modes of plant operation, make any physical changes, or 
alter any operational setpoints. Therefore the changes do not create 
the possibility of a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The relocated requirements do not satisfy 
the 10 CFR 50.36 criteria for inclusion in the Technical 
Specifications. Therefore, the changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Susan Frant Shankman, Acting.

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

    Date of amendment request: March 27, 1996.
    Description of amendment request: The amendment proposes to revise 
the Technical Specifications to support adoption of the primary 
containment leakage rate testing requirements of Option B to 10 CFR 50, 
Appendix J at the FitzPatrick plant, and clarify the numerical value of 
the allowable containment leakage rate (La) as 1.5 percent per 
day.
    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 Authority has evaluated the proposed TS Amendment and 
determined that it does not represent a significant hazards 
consideration. Based on the criteria for defining a significant 
hazards consideration established in 10 CFR 50.92, operation of the 
James A. FitzPatrick Nuclear Power Plant in accordance with the 
proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed changes do not involve a change to the design or 
operation of the plant. The systems affected by this proposed TS 
change are not assumed in any safety analyses to initiate any 
accident sequence. Therefore, the probability of any accident 
previously evaluated is not increased by this proposed TS change. 
The clarification of the allowable containment leakage rate 
(La) is consistent with the accident analyses. There is no 
change to the consequences of an accident previously evaluated 
because maintaining leakage within limits assumed in the accident 
analyses ensures that the dose consequences resulting from an 
accident are not increased. The proposed TS changes maintain an 
equivalent level of reliability

[[Page 20856]]

and availability for all affected systems. The ability of the 
affected systems associated with maintaining leak rate integrity to 
perform their intended function is unaffected by the proposed TS 
changes. Implementation of these changes will provide continued 
assurance that specified parameters associated with containment 
integrity will remain within acceptance limits, and as such, will 
not significantly increase the consequences of a previously 
evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because: 
    The proposed changes allow adoption of those requirements 
specified in Option B to 10 CFR 50, Appendix J, and do not involve a 
change to the plant design and operation. As a result, the proposed 
changes do not affect the parameters or conditions that could 
contribute to the initiation of any accidents. The methods of 
performing primary containment leakage rate testing are not changed. 
No new accident modes are created by allowing extended intervals for 
Type A, B and C testing, or by clarifying the numerical value of the 
allowable containment leakage rate (La). No safety-related 
equipment or safety functions are altered, or adversely affected, as 
a result of these changes. The proposed changes will not introduce 
failure mechanisms beyond those already considered in the current 
plant safety analyses. Extension of the test intervals, and 
clarification of the allowable leakage rate, does not contribute to 
the possibility of a new or different kind of accident or 
malfunction from those previously analyzed.
    3. Involve a significant reduction in the margin of safety 
because: The proposed changes affect the frequency of primary 
containment leakage rate testing, and the numerical definition of 
the allowable containment leakage rate (La). The design of the 
FitzPatrick plant is not changed. The methodology for test 
performance is unchanged and Type A, B and C tests will continue to 
be performed at Pa. The proposed changes provide 
sufficient controls to ensure that proper maintenance and repairs 
are performed on the primary containment, and systems and components 
penetrating the primary containment. The reliability of containment 
systems assumed to operate in the plant safety analyses is not 
reduced. The numerical value of La specified in Specification 
6.20 is consistent with the accident analyses, therefore, the dose 
consequences of any analyzed accidents are not increased. Therefore, 
the proposed changes provide continued assurance of the leak 
tightness of the containment without adversely affecting the public 
health and safety and, as such, 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 request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Susan Frant Shankman, Acting.

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

    Date of amendment request: April 22, 1996.
    Description of amendment request: The amendments would change the 
Technical Specifications to implement 10 CFR Part 50, Appendix J, 
Option B, for the Type A test by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leakage-Test Program.''
    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.
    Containment leak rate testing is not an initiator of any 
accident. The proposed changes do not make any physical changes to 
the containment. The proposed changes do not affect performance of 
the containment, reactor operations or accident analysis. Therefore, 
the proposed changes will not involve an increase in the probability 
of any previously evaluated accident.
    Since the allowable leakage rate is not being changed and since 
the analysis documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program'' concludes that the impact on public 
health and safety due to extended intervals is negligible, the 
proposed changes will not involve an increase in the consequences of 
any previously evaluated accident. Therefore, adoption of a 
performance-based verification of leakage rates for the overall 
containment boundary will provide an equivalent level of safety and 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change makes no physical changes to the plant. 
Since no physical changes are involved and since the analysis 
documented in NUREG-1493 confirms that the performance based 
schedule continues to maintain a minimal impact on public risk, it 
can be concluded that the effect of the containment on any accident 
will not change. The proposed change does not affect normal plant 
operations or configuration, nor does it affect leak rate test 
pressure.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes are based on NRC-accepted provisions, and 
maintain necessary levels of reliability of containment integrity. 
The performance-based approach to leakage rate testing recognizes 
that historically good results of containment testing provide 
appropriate assurance of future containment integrity. This supports 
the conclusion that the impact on the health and safety of the 
public as a result of extended test intervals is negligible. Since 
the analysis documented in NUREG-1493 confirms that the performance 
based schedule continues to maintain a minimal impact on public 
risk, it can be concluded that the margin of safety is not 
significantly affected by the proposed changes.
    The test history at Salem Units 1 and 2 (no ILRT failures) 
provides continued assurance of the leak tightness of the 
containment structure.
    Therefore, the proposed amendment will not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: April 4, 1996 (TS 96-01).
    Description of amendment request: The proposed change would revise 
the appropriate technical specifications, surveillances, and bases as 
needed for the conversion from Westinghouse nuclear fuel to Framatome 
Cogema Mark-BW17 nuclear fuel.
    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:


[[Page 20857]]


    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The analyses provided in Topical Report BAW-10220P show that the 
changes do not significantly change the results of previously 
evaluated events. These analyses provide the template for accident 
analyses assumptions that must be met by the cycle-specific reload 
analysis.
    The SQN Units 1 and 2 Cycle 9 reload cores with Mark-BW fuel 
will be designed to operate within the approved limits for accident 
analysis. The limits provided in the TS and described in the Updated 
Final Safety Analysis Report (UFSAR) provide the framework for 
accident analyses. By maintaining these limits, the probability or 
consequences of accidents related to the core changes do not 
significantly change. Thus, it is concluded that there is no 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The change to Mark-BW fuel cores and mixed (transition) cores 
has been evaluated in the Topical Report BAW-10220P. It was 
concluded that the change did not create new or different kinds of 
accidents. The change in fuel suppliers has been evaluated for 
consideration of the effects of power distribution and peaking 
factors such that there are no restrictions on the use of Mark-BW 
fuel assemblies beyond those already established in the UFSAR and 
TS. Adherence to the safety analysis limits restricts the 
possibility of new or different accidents. Historically, new 
accidents have not been associated with changes in fuel suppliers as 
long as safety analysis limits continue to be met. It is concluded 
that transition to Mark-BW fuel does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety is established by the acceptance criteria 
used by NRC. Meeting the acceptance criteria assures that the 
consequences of accidents are within known and acceptable limits. 
The loss-of-coolant accident (LOCA) acceptance criteria are 
unchanged: peak cladding temperature of 2200 degrees 
Fahrenheit, peak cladding oxidation of 17 percent, 
average clad oxidation of 1 percent, and long-term 
coolability. These requirements continue to be met. The methods used 
to demonstrate conformance with these limits have changed, and were 
reviewed to assure that the methods, as well as the results, are 
acceptable. The acceptance criteria for Departure from Nucleate 
Boiling (DNB) events has not changed and is still the 95 percent 
probability and 95 percent confidence interval that DNB is not 
occurring during the transient. The DNB correlation, and methods 
used to demonstrate that DNB limits are met, have changed, and these 
changes were reviewed to assure conformance with acceptable 
practices. Other changes, as well as the changes discussed above, 
have been evaluated in the referenced safety analyses and are shown 
to meet applicable acceptance criteria. Other margins, such as 
avoiding fuel centerline melting are not significantly changed. 
Based on these results, it is concluded that the margin of safety is 
not significantly reduced.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CPR 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, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of application request: April 12, 1996.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.4 and its associated Bases to 
address the installation of laser welded tube sleeves in the Callaway 
Plant steam generators.
    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 elevated tubesheet LWS [laser welded sleeve] configuration 
has been designed and analyzed in accordance with the requirements 
of the ASME [American Society of Mechanical Engineers] Code. The 
applied stresses and fatigue usage for the sleeve and weld are 
bounded by the limits established in the ASME Code. ASME Code 
minimum material property values are used for the structural and 
plugging limit analysis. Ultrasonic inspection is used to verify 
that minimum weld fusion zone thickness are produced. Mechanical 
testing has shown that the individual joint structural strength of 
Alloy 690 LWS under normal, upset and faulted conditions provides 
margin to the acceptance limits. These acceptance limits bound the 
most limiting (3 times normal operating pressure differential) burst 
margin recommended by RG [Regulatory Guide] 1.121. Therefore, each 
individual joint provides for structural integrity exceeding RG 
recommendations.
    Leakage testing for \7/8\'' and \3/4\'' tube sleeves has 
demonstrated that no unacceptable levels of primary to secondary 
leakage are expected during any plant condition, including the case 
where the seal weld is not produced in the lower joint of the 
tubesheet sleeve. Similar tests of 11/16'' tube sleeves will be 
completed prior to Refuel 8.
    The sleeve minimum acceptable wall thickness (used for 
developing the depth-based plugging limit for the sleeve) is 
determined using the guidance of Regulatory Guide 1.121 and the 
pressure stress equation of Section III of the ASME Code. The 
limiting requirement of Regulatory Guide 1.121, which applies to 
part throughwall degradation, is that the minimum acceptable wall 
must maintain a factor of safety of three against tube failure under 
normal operating (design) conditions. A bounding set of design and 
transient loading input conditions was used for the minimum wall 
thickness evaluation in the generic evaluation. Evaluation of the 
minimum acceptable wall thickness for normal, upset and postulated 
accident condition loading per the ASME Code indicates these 
conditions are bounded by the design condition requirement minimum 
wall thickness.
    A bounding tube wall degradation growth rate per cycle and an 
eddy current uncertainty has been assumed for determining the sleeve 
TS plugging limit. The sleeve wall degradation extent determined by 
eddy current examination, which would require plugging sleeved 
tubes, is developed using the guidance of RG 1.121 and is defined in 
WCAP-14596 to be 39 percent throughwall of the sleeve nominal wall 
thickness.
    The consequences of failure of the sleeve joint are bounded by 
the current steam generator tube rupture analysis included in the 
Callaway FSAR. Due to the slight reduction in diameter caused by the 
sleeve wall thickness, primary coolant release rates would be 
slightly less than assumed for the steam generator tube rupture 
analysis (depending on the break location), and therefore, would 
result in lower total primary fluid mass release to the secondary 
system.
    The proposed change does not adversely impact any other 
previously evaluated design basis accident of the results of LOCA 
and non-LOCA accident analyses for the current TS minimum reactor 
coolant system flow rate. The results of the analyses and testing 
demonstrate that the sleeve assembly is an acceptable means of 
maintaining tube integrity. Furthermore, per Regulatory Guide 1.83, 
``Inservice Inspection of Pressurized Water Reactor Steam Generator 
Tubes'' recommendations, the sleeved tube can be monitored through 
periodic inspections with present eddy current techniques. These 
measures demonstrate that installation of sleeves spanning degraded 
areas of the tube will restore the tube to a condition consistent 
with its original design basis.
    Corrosion testing of laser welded sleeve joints indicates that 
the corrosion resistance

[[Page 20858]]

(relative to roll transition control samples) can be increased by 
greater than a factor of ten with the application of a post weld 
heat treatment [PWHT]. All free span laser welds will receive a post 
weld heat treatment. Therefore, rapid corrosion degradation of the 
free span laser weld joint region is not expected. Recently 
performed corrosion testing of LWS joints in locked (at the first 
TSP [tube support plate] structure) tube conditions indicates that 
the PWHT, the stress corrosion cracking initiation potential in the 
weld region of the parent tube is reduced and the cracking 
resistance is enhanced. Similar test results and conclusions would 
be expected for Callaway based on the similarity of designs and 
expected tube far field residual stresses.
    Conformance of the sleeve design with the applicable sections of 
the ASME Code and results of the leakage and mechanical tests, 
support the conclusion that installation of LWS will not increase 
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.
    Sleeving will not adversely affect any plant component. Stress 
and fatigue analysis of the repair has shown that the ASME Code and 
Regulatory Guide 1.121 criteria are not exceeded. Implementation of 
LWS maintains overall tube bundle structural and leakage integrity 
at a level consistent to that of the originally supplied tubing 
during all plant conditions. Leak and mechanical testing of sleeves 
support the conclusions of the calculations that each sleeve joint 
retains both structural and leakage integrity during all conditions. 
Sleeving of tubes does not provide a mechanism resulting in an 
accident outside of the area affected by the sleeves. Any accident 
as a result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing tube rupture accident 
analysis.
    Implementation of LWS will reduce the potential for primary to 
secondary leakage during a postulated steam line break while not 
significantly impacting available primary coolant flow area in the 
event of a LOCA. By effectively isolating degraded areas of the tube 
through repair, the potential for steam line break leakage is 
reduced. These degraded intersections now are returned to a 
condition consistent with the Design Basis. While the installation 
of a sleeve reduces primary coolant flow, the reduction is far below 
that caused by plugging. Therefore, far greater primary coolant flow 
area is maintained through sleeving versus plugging.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The LWS repair of degraded steam generator tubes has been shown 
by analysis to restore the integrity of the tube bundle consistent 
with its original design basis condition, i.e., tube/sleeve 
operational and faulted condition stresses are bounded by the ASME 
Code requirements and the repaired tubes are leaktight. The safety 
factors used in the design of sleeves for the repair of degraded 
tubes are consistent with the safety factors in the ASME Code used 
in steam generator design. The design of the tubesheet sleeve lower 
joints for the \3/4\'' and \7/8\'' sleeves have been verified by 
testing to preclude leakage during normal and postulated accident 
conditions. Similar tests of \11/16\'' sleeves will be completed 
prior to Refuel 8. The portions of the installed sleeve assembly 
which represent the reactor coolant pressure boundary can be 
monitored for the initiation and progression of sleeve/tube wall 
degradation, thus satisfying the requirements of Regulatory Guide 
1.83. The portion of the tube bridged by the sleeve joints is 
effectively removed from the pressure boundary, and the sleeve then 
forms the new pressure boundary. The areas of the sleeved tube 
assembly which require inspection are defined in WCAP-14596.
    In addition, since the installed sleeve represents a portion of 
the pressure boundary, a baseline inspection of these areas is 
required prior to operation with sleeves installed. The effect of 
sleeving on the design transients and accident analyses has been 
reviewed based on the installation of sleeves up to the level of 
steam generator tube plugging coincident with the minimum reactor 
flow rate and the Callaway Safety Analysis.
    Provisional requirements cited in other NRC Safety Evaluation 
Reports addressing the implementation of sleeving have required the 
reduction of the individual steam generator normal operation primary 
to secondary leakage limit from 500 to 150 gpd. Consistent with 
these evaluations, Union Electric will reduce the per steam 
generator leak rate limit of 500 gpd in TS 3.4.6.2.c to 150 gpd. The 
establishment of this leakage limit at 150 gpd provides additional 
safety margin.
    Finally, Union Electric will reduce the tube plugging limit from 
48 percent through wall to 40 percent through wall to be consistent 
with NUREG-1431. The establishment of the plugging limit at 40 
percent through wall provides additional safety margin.

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

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

    Date of application request: April 12, 1996.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.4 and its associated Bases to 
address the installation of electrosleeves in the Callaway Plant steam 
generators.
    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 electrosleeve configuration has been designed and analyzed 
in accordance with the requirements of the ASME [American Society of 
Mechanical Engineers] Code. The applied stresses and fatigue usage 
for the sleeve are bounded by the limits established in the ASME 
Code. ASME Code minimum material property values are used for the 
structural and plugging limit analysis. Mechanical testing has shown 
that the structural strength of nickel electrosleeves under normal, 
upset and faulted conditions provides margin to the acceptance 
limits. These acceptance limits bound the most limiting (3 times 
normal operating pressure differential) burst margin recommended by 
RG [Regulatory Guide] 1.121. Leakage testing for \5/8\'', \7/8\'' 
and \3/4\'' tube sleeves has demonstrated that no unacceptable 
levels of primary to secondary leakage are expected during any plant 
condition. Similar tests of \11/16\'' tube electrosleeves will be 
completed prior to Refuel 8.
    The sleeve nominal wall thickness (used for developing the 
depth-based plugging limit for the sleeve) is determined using the 
guidance of Regulatory Guide 1.121 and the pressure stress equation 
of Section III of the ASME Code. The limiting requirement of 
Regulatory Guide 1.121, which applies to part throughwall 
degradation, is that the minimum acceptable wall must maintain a 
factor of safety of three against tube failure under normal 
operating (design) conditions. A bounding set of design and 
transient loading input conditions was used for the minimum wall 
thickness evaluation in the generic evaluation. Evaluation of the 
minimum acceptable wall thickness for normal, upset and postulated 
accident condition loading per the ASME Code indicates these 
conditions are bounded by the design condition requirement minimum 
wall thickness.
    A bounding tube wall degradation growth rate per cycle and an 
NDE [nondestructive examination] uncertainty has been assumed for 
determining the sleeve TS plugging limit. The sleeve wall 
degradation extent determined by NDE, which would require plugging 
sleeved tubes, is developed using the guidance of RG 1.121 and is 
defined in BAW-10219P to be 20 percent throughwall.
    The consequences of failure of the sleeve are bounded by the 
current steam generator tube rupture analysis included in the 
Callaway FSAR [final safety analysis report]. Due to the slight 
reduction in diameter caused by the sleeve wall thickness, primary 
coolant release rates would be slightly less

[[Page 20859]]

than assumed for the steam generator tube rupture analysis 
(depending on the break location), and therefore, would result in 
lower total primary fluid mass release to the secondary system.
    The proposed change does not adversely impact any other 
previously evaluated design basis accident or the results of LOCA 
[loss-of-coolant accident] and non-LOCA accident analyses for the 
current TS minimum reactor coolant system flow rate. The results of 
the analyses and testing demonstrate that the electrosleeve is an 
acceptable means of maintaining tube integrity. Furthermore, per 
Regulatory Guide 1.83 recommendations, the sleeved tube can be 
monitored through periodic inspections with present NDE techniques. 
These measures demonstrate that installation of sleeves spanning 
degraded areas of the tube will restore the tube to a condition 
consistent with its original design basis.
    Conformance of the electrosleeve design with the applicable 
sections of the ASME Code and results of the leakage and mechanical 
tests, support the conclusion that installation of electrosleeves 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Electrosleeving does not represent a potential to adversely 
affect any plant component. Stress and fatigue analysis of the 
repair has shown that the ASME Code and Regulatory Guide 1.121 
criteria are not exceeded. Implementation of electrosleeving 
maintains overall tube bundle structural and leakage integrity at a 
level consistent to that of the originally supplied tubing during 
all plant conditions. Leak and mechanical testing of electrosleeves 
support the conclusions of the calculations that each sleeve retains 
both structural and leakage integrity during all conditions. 
Sleeving of tubes does not provide a mechanism resulting in an 
accident outside of the area affected by the sleeves. Any accident 
as a result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing tube rupture accident 
analysis.
    Implementation of sleeving will reduce the potential for primary 
to secondary leakage during a postulated steam line break while not 
significantly impacting available primary coolant flow area in the 
event of a LOCA. By effectively isolating degraded areas of the tube 
through repair, the potential for steam line break leakage is 
reduced. These degraded intersections now are returned to a 
condition consistent with the Design Basis. While the installation 
of a sleeve reduces primary coolant flow, the reduction is far below 
that caused by plugging. Therefore, far greater primary coolant flow 
area is maintained through sleeving versus plugging.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The electrosleeve repair of degraded steam generator tubes has 
been shown by analysis to restore the integrity of the tube bundle 
consistent with its original design basis condition, i.e., tube/
sleeve operational and faulted condition stresses are bounded by the 
ASME Code requirements and the repaired tubes are leaktight. The 
safety factors used in the design of sleeves for the repair of 
degraded tubes are consistent with the safety factors in the ASME 
Code used in steam generator design. The portions of the installed 
sleeve assembly which represent the reactor coolant pressure 
boundary can be monitored for the initiation and progression of 
sleeve/tube wall degradation, thus satisfying the requirements of 
Regulatory Guide 1.83. The portion of the tube bridged by the sleeve 
is effectively removed from the pressure boundary, and the sleeve 
then forms the new pressure boundary. The areas of the sleeved tube 
assembly which require inspection are defined in BAW-10219P.
    In addition, since the installed sleeve represents a portion of 
the pressure boundary, a baseline inspection of these areas is 
required prior to operation with sleeves installed. The effect of 
sleeving on the design transients and accident analyses has been 
reviewed based on the installation of sleeves up to the level of 
steam generator tube plugging coincident with the minimum reactor 
flow rate and the Callaway Safety Analysis.
    Provisional requirements cited in other NRC Safety Evaluation 
Reports addressing the implementation of sleeving have required the 
reduction of the individual steam generator normal operation primary 
to secondary leakage limit from 500 to 150 gpd.
    Consistent with these evaluations, Union Electric will reduce 
the per steam generator leak rate limit of 500 gpd in TS 3.4.6.2.c 
to 150 gpd. The establishment of this leakage limit at 150 gpd 
provides additional safety margin.
    Finally, Union Electric will reduce the tube plugging limit from 
48 percent through wall to 40 percent through wall to be consistent 
with NUREG-1431. The establishment of the plugging limit at 40 
percent through wall provides additional safety margin.

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

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: April 4, 1996.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications regarding secondary containment 
integrity including addition of required actions in the event secondary 
containment integrity is not maintained when required. It would also 
require surveillance of the secondary containment isolation valves 
under the licensee's in-service testing program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed changes do not result in any hardware changes. 
The requirements for Secondary Containment integrity are not assumed 
in the initiation of any analyzed event. The proposed changes 
establish and maintain adequate assurance that Secondary Containment 
Integrity will be maintained as assumed in analyses for the 
mitigation of accident consequences. Not requiring Secondary 
Containment Integrity when the reactor coolant system is not vented 
in the Cold Shutdown condition or the Refuel Mode does not involve 
an increase in previously evaluated accident consequences since no 
mechanism exists to impart additional fission-products into the 
reactor coolant. Under these conditions, activities for which the 
reactor coolant system would not be vented would be strictly 
controlled and monitored. As a result, leaks or pipe breaks would 
typically be detected before significant inventory loss occurred. 
These activities would typically be performed after refueling when 
few noncondensible gases remain in the reactor coolant. The 
temperature limitation of 212 deg.F will ensure that water, not 
steam, would be emitted from the postulated leak or pipe break. In 
addition, under these conditions, stored energy is sufficiently low 
that even with loss of inventory following a recirculation line 
break, core coverage would be maintained by the low pressure 
emergency core cooling systems required per Specification 3.5.H and 
the fuel would not exceed its peak clad temperature limit. As a 
result, the potential for failed fuel and a subsequent increase in 
reactor coolant activity is minimized and significant releases of 
radioactive material to the environment would not be expected to 
occur. Therefore, these changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) The proposed changes do not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in parameters governing normal operation and will not 
alter the method used by any system to perform its design function. 
The proposed changes to not allow plant operation in any mode that 
is not already evaluated and will still ensure Secondary Containment 
Integrity is maintained when required. Thus, these changes do not 
create the possibility of a new or different kind of

[[Page 20860]]

accident from any accident previously evaluated.
    (3) The proposed changes to Secondary Containment Integrity 
requirements have no impact on any safety analysis assumptions. 
Secondary Containment Integrity will be maintained as assumed in the 
safety analyses and as stated in current Bases 3.7.B and 3.7.C. Not 
requiring Secondary Containment Integrity when the reactor coolant 
system is not vented in the Cold Shutdown condition or the Refuel 
Mode does not involve significant reduction in a margin of safety 
since no mechanism exists to impart additional fission products into 
the reactor coolant. Under these conditions, activities for which 
the reactor coolant system would not be vented would be strictly 
controlled and monitored. As a result, leaks or pipe breaks would 
typically be detected before significant inventory loss occurred. 
These activities would typically be performed after refueling, at 
low decay levels, and with reactor coolant temperature less than or 
equal to 212 deg.F. In addition, under these conditions, stored 
energy in the reactor core is very low. The reactor pressure vessel 
would rapidly depressurize in the event of a large primary system 
leak and the low pressure emergency core cooling systems required 
per Specification 3.5.H under these conditions would be adequate to 
keep the core flooded. This would ensure that the fuel would not be 
uncovered and would not exceed the peak clad temperature limit.
    As a result, the potential for failed fuel and a subsequent 
increase in reactor coolant activity is minimized and significant 
releases of radioactive material to the environment would not be 
expected to occur. Therefore, these changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: R.K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624.
    NRC Project Director: Susan Frant Shankman.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: April 4, 1996.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements for control rod over-travel to 
remove the specific testing methodology from the Technical 
Specifications to administratively controlled documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The control rod drive mechanism over-travel is not 
considered to be the initiator of any previously analyzed accident. 
Verification of coupling of the control rods and drive mechanisms is 
performed by other means and continues to be required in the same 
manner, so there is no significant increase in the probability of a 
rod drop accident. The over-travel indication is also not considered 
in the mitigation of consequences of any previously analyzed 
accident, and the removal of a specific surveillance of the 
indication will not affect the response of the control rods or the 
reactor protection system to these accidents. Therefore, this change 
will not significantly increase the probability or consequences of 
any previously analyzed accident.
    (2) The proposed change does not necessitate a physical 
alteration of the plant (no new or different type of equipment will 
be installed) nor changes in parameters governing normal plant 
operation. The proposed change will continue to provide effective 
methods to assure the control rods and their drive mechanisms are 
coupled and preserve the safety functions associated with the 
prevention or automatic mitigation of design basis accidents. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) The proposed changes continue to provide an appropriate 
method for verification of the capability of the over-travel 
indication to perform its function. Therefore, this change will not 
significantly reduce a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: R.K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624.
    NRC Project Director: Susan Frant Shankman.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: April 15, 1996.
    Description of amendment request: The proposed changes will clarify 
the applicability of the quadrant power tilt ratio (QPTR) requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Operation of Surry Power Station in accordance with the proposed 
Technical Specifications change will not:
    1. Involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated.
    The application of the QPTR limits, as proposed, will assure 
that the gross core radial power distribution remains consistent 
with design limits above 50% power. At or below 50% rated thermal 
power, there is insufficient stored energy in the fuel or 
insufficient energy being transferred to the reactor coolant to 
require implementation of a QPTR limit on the distribution of core 
power. Therefore, the proposed change to clarify the applicability 
of the QPTR requirements has no impact on the probability of an 
accident occurrence and does not increase the consequences of any 
design basis accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no plant modifications or changes in methods of plant 
operation introduced by the proposed change. The change would limit 
the application of QPTR limits to operation at power levels >50% to 
preclude core power distributions from occurring which would violate 
fuel design criteria previously analyzed. At or below 50% rated 
thermal power, there is no impact to core power distributions which 
could affect the fuel design criteria. Therefore, the proposed 
change does not create the possibility for an accident or 
malfunction of a different type than that previously evaluated in 
the safety analysis report.
    3. Involve a significant reduction in a margin of safety.
    The proposed change only affects the applicability of the QPTR 
limits. The QPTR limits remain unchanged to preclude any violation 
of previously analyzed fuel design criteria. Adherence to the QPTR 
limits, hot channel factors, and applicable Limiting Conditions for 
Operation will continue. Therefore, the margin of safety as 
described in the Bases Section of any part of the Technical 
Specifications is not reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams,

[[Page 20861]]

Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Project Director: Eugene V. Imbro.

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.

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

    Date of amendment request: April 3, 1996.
    Description of amendment request: The proposed amendments would 
revise the hydrogen mitigation system Technical Specifications (TS). 
The change would provide that, if neither the Train A or Train B 
igniter is operable in any one containment region, then there is an 
allowance of 7 days to restore one hydrogen igniter to OPERABLE status, 
or be in Hot Shutdown within the next 6 hours. This would be consistent 
with the guidance of the Standard TS for Westinghouse plants, NUREG-
0431.
    Date of publication of individual notice in Federal Register: April 
16, 1996 (61 FR 16649).
    Expiration date of individual notice: May 16, 1996.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, Docket 
No. 50-498, South Texas Project, Unit 1, Matagorda County, Texas

    Date of amendment request: January 22, 1996, as supplement by 
letter dated April 4, 1996.
    Brief description of amendments: The proposed amendment would 
modify the steam generator tube plugging criteria in Technical 
Specification 3/4.4.5, Steam Generators, and the allowable leakage in 
Technical Specification 3/4.4.6.2, Operational Leakage, and the 
associated Bases. The proposed amendment would allow the implementation 
of steam generator voltage-based repair criteria for the tube support 
plate (TSP)/tube intersections for Unit 1.
    Date of individual notice in the Federal Register: April 16, 1996 
(61 FR 16651)
    Expiration date of individual notice: May 16, 1996.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

    Date of amendment request: March 14, 1996.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications for Indian Point Nuclear Generating 
Unit No. 3 to allow a one-time extension of the test intervals for the 
pressurizer safety valve setpoint and snubber functional testing that 
is due in May 1996.
    Date of publication of individual notice in Federal Register: April 
3, 1996 (61 FR 14835)
    Expiration date of individual notice: May 3, 1996.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.

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.

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 delete from the 
Technical Specifications certain review responsibilities of the Plant 
Operations Review Committee and the Site Operations Review Committee 
relating to the Emergency Plan and the Security Plan and their 
respective implementing procedures. The proposed changes are consistent 
with the guidance of Generic Letter 93-07.
    Date of issuance: April 24, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 189, 94, 196, and 128
    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 14, 1996 (61 
FR 5812)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 24, 1996.
    No significant hazards consideration comments received: No.

[[Page 20862]]

    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, CT 06360, and the Waterford 
Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, CT 06385, 
for Millstone 1, 2, and 3.

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

    Date of application for amendments: January 11, 1996, as 
supplemented by letter dated April 2, 1996.
    Brief description of amendments: The amendments revise Technical 
Specification Table 3.6-1, Table 3.6-2a and Table 3.6-2b to delete 
references to process penetration M308 and service water system (RN) 
valves RN-429A and RN-432B from the lists of secondary containment 
bypass valves and containment isolation valves. The RN valves are no 
longer in service and are planned to be removed in forthcoming outages. 
The penetration will then be capped with blank flanges.
    Date of issuance: April 23, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 143 and 137
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5813) The April 2, 1996, letter provided additional information that 
did not change the scope of the January 11, 1996, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated April 23, 1996.
    No significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

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

    Date of application of amendments: December 7, 1995.
    Brief description of amendments: The amendments revise Secondary 
Decay Heat Removal Technical Specification (TS) 3.4.2 and TS Table 4.1-
1 to delete the requirement of having the main feedwater pump discharge 
header pressure switch provide an input to actuate the Anticipatory 
Reactor Trip System and Emergency Feedwater System.
    Date of Issuance: April 15, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 216, 216, 213.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1628).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 15, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

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: October 25, 1993, as 
supplemented August 31, 1994, and October 5, 1995.
    Brief description of amendments: The amendments modify the 
surveillance requirements related to dune survey and mangrove swamp 
monitoring and relocate them to the Final Safety Analysis Report
    Date of Issuance: April 11, 1996.
    Effective Date: April 11, 1996.
    Amendment Nos.: 142 and 82.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67844) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 11, 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

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

    Date of application for amendment: August 31, 1995, as supplemented 
February 29, 1996.
    Brief description of amendment: The amendment revises License 
Condition 2.B(6)(c), Fire Protection, and relocates fire protection 
requirements from the Maine Yankee Atomic Power Station Technical 
Specifications to the Maine Yankee Fire Protection Plan. The amendment 
is consistent with the guidance of NRC Generic Letters 86-10, 
Implementation of Fire Protection Requirements, and 88-12, Removal of 
Fire Protection Requirements, from the Technical Specifications.
    Date of issuance: April 5, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 156.
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52932) The February 29, 1996, letter provided document dates that did 
not change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 5, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

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

    Date of application for amendment: October 25, 1995.
    Brief description of amendment: The amendment changes the Technical 
Specification regarding the average power range monitor (APRM) 
setpoints. These changes establish limiting conditions for operations 
and surveillance requirements for the APRM flow-biased scram and rod 
block setpoints. The amendment also incorporates several editorial 
changes and renumbered pages, removal of blank pages, revised Table of 
Contents, and modified Bases section for APRM setpoint requirements.
    Date of issuance: April 15, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 93.
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65682).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1996.
    No significant hazards consideration comments received: No.

[[Page 20863]]

    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360 and at the temporary local public document room 
located at the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry 
Road, Waterford, CT 06385.

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

    Date of amendment request: August 4, 1995, as supplemented by 
letter dated January 22, 1996.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TS) for the requirements for the containment 
radiation high signal (CRHS) and the safety injection and refueling 
water (SIRW) tank low signal (STLS) contained in TS 2.15, Tables 2-3 
and 2-4. Table 3-2 of TS 3.1 will also be revised to include 
administrative changes to the CRHS surveillance methods to be 
consistent with the applicable surveillance functions. The Basis of TS 
2.15 is being revised to clarify that the number of installed channels 
for CRHS is two. The term ``SOURCE CHECK'' is being deleted from the 
Definitions section.
    Date of issuance: April 24, 1996.
    Effective date: April 24, 1996.
    Amendment No.: 173.
    Facility Operating License No. DPR-40. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45182).
    The January 22, 1996, supplemental letter provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 24, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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

    Date of application for amendments: March 1, 1995, as supplemented 
by letter dated April 16, 1996.
    Brief description of amendments: The amendments change the 
concentration of calibration gas required to calibrate the Hydrogen and 
Oxygen Analyzers, and support the requirements of Limerick Generating 
Station Transient Response Implementation Plan (TRIP) T-102, ``Primary 
Containment Control Bases.''
    Date of issuance: April 23, 1996.
    Effective date: Both units, as of date of issuance, to be 
implemented within 45 days.
    Amendment Nos.: 116 and 78.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20525) The April 16, 1996 letter requested a new effective date and did 
not change the initial proposed no significant hazards consideration 
determination nor the Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 23, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    Date of application for amendments: August 21, 1992; supplemented 
September 3, 1993, and March 28, 1996 (TS 92-07).
    Brief description of amendments: The amendments revise the 
allowable value for the reactor coolant system loss of flow reactor 
trip setpoint from greater than or equal to 89.4 percent to greater 
than or equal to 89.6 percent.
    Date of issuance: April 26, 1996.
    Effective date: April 26, 1996.
    Amendment Nos.: 221 and 212.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: September 30, 1992 (57 
FR 45090). The September 3, 1993 and March 28, 1996 supplemental 
letters provided clarifying information which did not change the 
proposed no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 26, 1996.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: February 5, 1996.
    Brief description of amendment: This amendment clarifies TS 3/
4.3.2.1, Table 3.3-3, Safety Features Actuation System Instrumentation, 
and revises Bases 3/4.3.1 and 3/4.3.2, Reactor Protection System and 
Safety System Instrumentation, to accurately reflect the design and 
actuation logic of the diesel generator load sequencer and the 
essential bus undervoltage relays.
    Date of issuance: April 23, 1996.
    Effective date: April 23, 1996.
    Amendment No.: 211.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10397).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of application for amendments: November 20, 1995, as 
supplemented March 14, 1996.
    Brief description of amendments: These amendments would permit the 
use of 10 CFR Part 50 Appendix J, Option B, performance-based 
containment leakage rate testing.
    Date of issuance: April 18, 1996.
    Effective date: April 18, 1996.
    Amendment Nos. 208 and 208.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65686) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 18, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

[[Page 20864]]

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 7, 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

[[Page 20865]]

must provide sufficient information to show that a genuine dispute 
exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, 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 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Tennesse Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Limestone County, Alabama

    Date of application for amendment: April 14, 1996.
    Brief description of amendment: The proposed amendment clarifies 
operability requirements for reactor vessel water level instrumentation 
to permit testing of components required by technical specifications.
    Date of issuance: April 16, 1996.
    Effective date: April 16, 1996.
    Amendment Nos.: 229, 244, and 204.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendment revises the technical specifications.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration, are contained in a Safety Evaluation dated April 
16, 1996. Public comments requested as to proposed no significant 
hazards consideration: No.
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: April 18, 1996.
    Brief description of amendment: The amendment approves the use of 
the station black out diesel generator in lieu of the emergency diesel 
generator associated with decay heat removal loop 2 during the tenth 
refueling outage. This condition will continue as long as no work is 
performed in the switchyard or on the SBODG or the remaining emergency 
diesel generator and a shutdown risk contingency plan is developed to 
ensure challenges to spent fuel pool cooling are minimized. This 
condition is expected to last for no more than seven days.
    Date of issuance: April 19, 1996.
    Effective date: April 19, 1996.
    Amendment No.: 210.
    Facility Operating License No. NPF-3: This amendment approved a 
one-time change to the design basis as described in the Updated Safety 
Analysis Report.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated April 
19, 1996.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus.

    Dated at Rockville, Maryland, this 1st of May 1996.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 96-11295 Filed 5-7-96; 8:45 am]
BILLING CODE 7590-01-P