[Federal Register Volume 61, Number 158 (Wednesday, August 14, 1996)]
[Notices]
[Pages 42274-42290]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10814]


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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 July 20, 1996, through August 2, 1996. The 
last biweekly notice was published on July 31, 1996 (61 FR 40013).

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

[[Page 42275]]

failure to act in a timely way would result, for example, in derating 
or shutdown of the facility, the Commission may issue the license 
amendment before the expiration of the 30-day notice period, provided 
that its final determination is that the amendment involves no 
significant hazards consideration. The final determination will 
consider all public and State comments received before action is taken. 
Should the Commission take this action, it will publish in the Federal 
Register a notice of issuance and provide for opportunity for a hearing 
after issuance. The Commission expects that the need to take this 
action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By September 13, 1996, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, 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

[[Page 42276]]

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: July 26, 1996
    Description of amendments request: The proposed amendment will 
revise the appropriate Technical Specifications and their Bases to 
permit the electrosleeving repair technique developed by Framatome 
Technologies, Inc. to be used at Calvert Cliffs Nuclear Power Plant 
(CCNPP). Electrosleeving is a steam generator tube repair method where 
an ultra-fine grained nickel is electrochemically deposited on the 
inner surface of a tube to form a structural repair of the degraded 
tube. The electrodeposition of nickel provides a continuous 
metallurgical bond that eliminates all leak paths and macro-crevices. 
The electroformed sleeve provides a structural, leak-tight seal, 
without deforming or changing the microstructure of the parent tube. 
Thus, unlike the conventional welded sleeves, electrosleeving does not 
require a post-installation stress relief.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The implementation of the proposed steam generator tube 
electrosleeving has been reviewed for impact on the current CCNPP 
licensing basis.
    Since the electrosleeve is designed using the applicable 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code as guidance, it meets the objectives of the original 
steam generator tubing. The applied stresses and fatigue usage for 
the electrosleeve are bounded by the limits established in the ASME 
Code. American Society of Mechanical Engineers 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 (three times normal operating pressure 
differential) burst margin recommended by Regulatory Guide 1.121. 
Burst testing of electrosleeved tubes has demonstrated that no 
unacceptable levels of primary-to-secondary leakage are expected 
during any plant condition.
    As in the original tube, the electrosleeve Technical 
Specification depth-based plugging limit is determined using the 
guidance of Regulatory Guide 1.121 and the pressure stress equation 
of Section III of the ASME Code. A bounding tube wall degradation 
growth rate per cycle and a nondestructive examination uncertainty 
has been assumed for determining the electrosleeve plugging limit.
    Evaluation of the proposed electrosleeved tubes indicates no 
detrimental effects on the electrosleeve or electrosleeve-tube 
assembly from reactor system flow, primary or secondary coolant 
chemistries, thermal conditions or transients, or pressure 
conditions as may be experienced at Calvert Cliffs. Corrosion 
testing of electrosleeve-tube assemblies indicates no evidence of 
electrosleeve or tube corrosion considered detrimental under 
anticipated service conditions.
    The implementation of the proposed electrosleeve has no 
significant effect on either the configuration of the plant, or the 
manner in which it is operated. The hypothetical consequences of 
failure of the electrosleeved tube is bounded by the current steam 
generator tube rupture analysis described in Section 14.15 of the 
Calvert Cliffs Updated 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 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.
    Therefore, BGE [Baltimore Gas and Electric] has concluded 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 kind 
of accident from any other accident previously evaluated.
    As discussed above, the electrosleeve is designed using the 
applicable ASME Code as guidance; therefore, it meets the objectives 
of the original steam generator tubing. As a result, the functions 
of the steam generators will not be significantly affected by the 
installation of the proposed electrosleeve. Adhesion and ductility 
tests performed per ASTM [American Society for Testing and 
Materials] standards verified that the electrosleeve will not fail 
by de-bonding or cracking. In addition, the proposed electrosleeve 
does not interact with any other plant systems. Any accident as a 
result of potential tube or electrosleeve degradation in the 
repaired portion of the tube is bounded by the existing tube rupture 
accident analysis. The continued integrity of the installed 
electrosleeve is periodically verified by the Technical 
Specification requirements.
    The implementation of the proposed electrosleeves has no 
significant effect on either the configuration of the plant, or the 
manner in which it is operated. Therefore, BGE concludes that this 
proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The repair of degraded steam generator tubes via the use of the 
proposed electrosleeve restores the structural integrity of the 
faulted tube under normal operating and postulated accident 
conditions. The design safety factors utilized for the electrosleeve 
are consistent with the safety factors in the ASME Boiler and 
Pressure Vessel Code used in the original steam generator design. 
The repair limit for the proposed electrosleeve is consistent with 
that established for the steam generator tubes. The portions of the 
installed electrosleeve assembly which represent the reactor coolant 
pressure boundary can be monitored for the initiation and 
progression of electrosleeve/tube wall degradation, thus satisfying 
the requirements of Regulatory Guide 1.83. Use of the previously 
identified design criteria and design verification testing assures 
that the margin to safety with respect to the implementation of the 
proposed electrosleeve is not significantly different from the 
original steam generator tubes.
    Therefore, BGE concludes that the proposed changes 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 
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: Jocelyn A. Mitchell, Acting Director

Carolina Power & Light Company, et al., Docket No. 50-325, 
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North 
Carolina

    Date of amendment request: April 8, 1996, as supplemented on July 
30, 1996. This notice supersedes the Federal Register notice published 
on June 5, 1996 (61 FR 28607).
    Description of amendment request: The licensee has proposed to 
revise the Technical Specifications (TS) to include the following 
changes: 1. The Minimum Critical Power Ratio (MCPR) Safety Limit 
specified in TS 2.1.2 from 1.07 to 1.10 for Unit 1 Cycle 11 operation; 
TS 5.3.1 to reflect the new fuel type (GE13) that will be inserted 
during Unit 1 Refueling Outage 10; 2. The acceptable range of sodium 
pentaborate concentration for the standby liquid control system shown 
in TS Figure

[[Page 42277]]

3.1.5-1 to reflect changes to poison material concentration needed to 
achieve reactor shutdown based on the new GE13 fuel type.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Proposed Change 1:
    The proposed license amendment will allow the loading and use of 
GE13 fuel assemblies in the Brunswick Unit 1 reactor core. The use 
of GE13 fuel assemblies requires that the safety limit minimum 
critical power ratio value also be revised. The safety limit minimum 
critical power ratio is established to maintain fuel cladding 
integrity during operational transients. The GE13 fuel assembly 
design has been analyzed using methods that have been previously 
approved by the Nuclear Regulatory Commission and documented in 
General Electric Nuclear Energ's reload licensing methodology 
Topical Report NEDE-24011, ``General Electric Standard Application 
for Reactor Fuel (GESTAR II).``Based on a cycle-specific calculation 
performed by General Electric, a safety limit minimum critical power 
ratio value of 1.10 has been established for the GE13 fuel type for 
Brunswick Unit 1 Cycle 11 operation. The cycle-specific calculation 
has been performed in accordance with the methodology in Revision 12 
of NEDE-24011. This cycle-specific calculation has demonstrated that 
a safety limit minimum critical power ratio value of 1.10 will 
ensure that 99.9 percent of the fuel rods avoid boiling transition 
during a transient event when all uncertainties are considered. The 
safety limit minimum critical power ratio value of 1.10 assures that 
fuel cladding protection equivalent to that provided with the 
existing safety limit minimum critical power ratio value is 
maintained. This ensures that the consequences of previously 
evaluated accidents are not significantly increased.
    The proposed revision of the safety limit minimum critical power 
ratio does not alter any plant safety-related equipment, safety 
function, or plant operations that could change the probability of 
an accident. The change does not affect the design, materials, or 
construction standards applicable to the fuel bundles in a manner 
that could change the probability of an accident.
    Proposed Change 2:
    The standby liquid control system provides a means of reactivity 
control that is independent of the normal reactivity control system. 
The standby liquid control system must be capable of assuring that 
the reactor core can be placed in a subcritical condition at any 
time during reactor core life. Technical Specification Figure 3.1.5-
1 specifies the acceptable range of concentrations and volumes for 
sodium pentaborate solution used as a neutron absorber (i.e., for 
reactivity control). The portion of the sodium pentaborate 
concentration range shown in Technical Specification Figure 3.1.5-1 
applicable to the lower range of tank volumes is being revised to 
increase the required concentration of sodium pentaborate solution. 
This change is needed to account for the additional shutdown 
reactivity needed based on the planned use of GE13 fuel assemblies 
as reload fuel for the Unit 1 reactor core. Since the standby liquid 
control system is independent from the normal means of controlling 
reactor core reactivity and not used to control core reactivity 
during normal plant operations, the proposed revision to the sodium 
pentaborate concentration curve for the standby liquid control 
system does not alter any plant safety-related equipment, safety 
function, or plant operations that could change the probability of 
an accident.
    The current volume-concentration range of sodium pentaborate 
used in the standby liquid control system will achieve a sufficient 
concentration of boron in the reactor vessel to ensure reactor 
shutdown. Based on the increased reactivity of the new GE13 reload 
fuel assemblies, the required sodium pentaborate volume-
concentration range is being revised to ensure sufficient neutron 
absorbing solution is available to achieve reactor shutdown; 
therefore, the consequences of an accident previously evaluated are 
not significantly increased.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Proposed Change 1:
    The GE13 fuel assembly has been designed and complies with the 
acceptance criteria contained in General Electric Nuclear Energy's 
standard application for reactor fuel (GESTAR-II), which provides 
the latest acceptance criteria for new General Electric fuel 
designs. The similarity of the GE13 fuel design to the previously 
accepted GE11 fuel design, in conjunction with the increased 
critical power capability of the GE13 fuel design, ensure that no 
new mode or condition of plant operation is being authorized by the 
loading and use of the GE13 fuel type. The proposed revision of the 
safety limit minimum critical power ratio from 1.07 to 1.10 does not 
modify any plant controls or equipment that will change the plant's 
responses to any accident or transient as given in any current 
analysis. Therefore, the proposed change to allow the loading and 
use of the GE13 fuel type and the revision of the safety limit 
minimum critical power ratio value from 1.07 to 1.10 will not create 
the possibility for a new or different kind of accident from any 
accident previously evaluated.
    Proposed Change 2:
    As discussed above, the standby liquid control system provides a 
means of reactivity control that is independent of the normal 
reactivity control system and is capable of assuring that the 
reactor core can be placed in a subcritical condition at any time 
during reactor core life. The proposed revision to the sodium 
pentaborate concentration range does not modify the standby liquid 
control system or its controls, does not modify other plant systems 
and equipment, and does not permit a new or different mode of plant 
operation. As such, the proposed revision to the minimum pentaborate 
concentration value does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    Proposed Change 1:
    As previously discussed, the GE13 fuel assembly design has been 
analyzed using methods that have been previously approved by the 
Nuclear Regulatory Commission and documented in General Electric 
Nuclear Energy's reload licensing methodology Topical Report NEDE-
24011, ``General Electric Standard Application for Reactor Fuel 
(GESTAR II).``The safety limit minimum critical power ratio value is 
selected to maintain the fuel cladding integrity safety limit (i.e., 
that 99.9 percent of all fuel rods in the core are expected to avoid 
boiling transition during operational transients). Appropriate 
operating limit minimum critical power ratio values are established, 
based on the safety limit minimum critical power ratio value, to 
ensure that the fuel cladding integrity safety limit is maintained. 
The operating limit minimum critical power ratio values are 
incorporated in the Core Operating limits Report as required by 
Technical Specification 6.9.3.1.
    Based on the cycle-specific calculation performed by General 
Electric, a safety limit minimum critical power ratio value of 1.10 
has been established for the GE13 fuel type for Unit 1 Cycle 11 
operation. This cycle-specific calculation has been performed based 
on the methodology contained in Revision 12 of NEDE-24011-P-A. The 
new GE13 safety limit minimum critical power ratio value of 1.10 for 
Unit 1 Cycle 11 operation is based on the same fuel cladding 
integrity safety limit criteria as that for the GE11 safety limit 
minimum critical power ratio (i.e., that 99.9 percent of all fuel 
rods in the core are expected to avoid boiling transition during 
operational transients); therefore, the proposed change does not 
result in a significant reduction in the margin of safety.
    Proposed Change 2:
    As previously stated, the purpose of the standby liquid control 
is to inject a neutron absorbing solution into the reactor in the 
event that a sufficient number of control rods cannot be inserted to 
maintain subcriticality. Sufficient solution is to be injected such 
that the reactor will be brought from maximum rated power conditions 
to subcritical over the entire reactor temperature range from 
maximum operating to cold shutdown conditions. General Electric 
methodology establishes a fuel type dependent standby liquid control 
system shutdown margin to account for calculational uncertainties. 
General Electric calculations show that an in-vessel concentration 
of 660 ppm will provide a standby liquid control system minimum 
shutdown margin in excess of the 3.2% delta k value required for the 
GE13 fuel. To achieve an in-vessel concentration of 660 ppm, the 
acceptable range of standby liquid control system tank 
concentrations is being

[[Page 42278]]

revised for the lower range of tank volumes. Thus, the proposed 
revision of the standby liquid control system sodium pentaborate 
volume-concentration range ensures that there will not be a 
significant reduction in the amount of available shutdown margin 
and, therefore, not 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    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: June 21, 1996
    Description of amendment request: The proposed amendments would 
extend the surveillance interval for TS 4.7.2.b and 4.7.2.d related to 
testing of the Control Room Emergency Filtration System from 18 months 
to 24 months. The amendments would also include a one-time extension of 
the allowed outage time for the Control Room and Auxiliary Electric 
Equipment Room Emergency Filtration System to allow each subsystem to 
be inoperable for up to 30 days during modifications to replace the 
existing deep bed charcoal absorbers with tray-type units.
    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:
    This Technical Specification change does not involve accident 
initiators or initial accident assumptions. The Control Room and 
Auxiliary Equipment Room Emergency Filtration System (CREFS) trains 
A and B are post-accident atmospheric cleanup components that are 
designed to limit the radiation exposure to personnel occupying the 
Control Room to 5 rem or less whole body during and following all 
design basis accident conditions. Therefore, this Technical 
Specification change does not increase the probability of occurrence 
of an accident previously evaluated.
    CREFS trains A and B are utilized to control the onsite dose to 
personnel in the Control Room. This Technical Specification change 
extends the [Limiting Condition for Operation] LCO duration for 
allowing each train to be inoperable one at a time from 7 days to 30 
days total for the current surveillance interval. This change is a 
one time change to allow for the repair/replacement work associated 
with the corroded filter unit charcoal retaining screens in the high 
efficiency charcoal adsorber section of each train. The...normal 
preventative maintenance and testing [will] be performed on the 
operable CREFS train just prior to taking the [opposite] filter 
train out of service for the modification. This action will ensure 
that the remaining subsystem is operable and ensure maximum 
reliability of the system. The Technical Specification change will 
not affect onsite dose if a [design-basis accident] DBA occurs and 
the operating filter unit does not fail. The operable filter unit 
will be sufficient to maintain the operating areas habitable. The 
original LCO allowed 7 day operation with only one operable train 
and is also susceptible to a single failure during the Allowed 
Outage Time. The probability that a DBA will occur coupled with the 
single failure of the operable train during the extended allowed 
outage time per the Technical Specification change is the same order 
of magnitude as for the current 7 day allowed outage time. 
Therefore, this change does not increase the consequences of an 
accident previously evaluated.
    The extension of the surveillance interval from 18 months to 24 
months extends the maximum interval between TS surveillances of the 
filter trains from 22.5 months to 30 months. The equipment that is 
affected are the CREFS filter trains A and B, which are comprised of 
HEPA filters, heaters, charcoal adsorbers, and fans. This equipment 
has a history of satisfactory surveillance testing (in-place testing 
and laboratory analysis of charcoal), and has had little maintenance 
problems for the past 5 years. Although the SER Section 6.4.1 and 
the [Regulatory Guide] RG 1.52 state that the units shall be tested 
every 18 months, a review of the basis documents for the testing 
(ANSI N510) shows that the 1975 edition recommended annual testing 
and later editions (1980 and 1989) state that testing be performed 
``at least once every operating cycle''. Therefore the extension of 
the surveillance intervals from 18 months to 24 months will not 
increase the consequences of an accident previously evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    This Technical Specification change will allow each train of 
CREFS to be inoperable one at a time for up to 30 days to repair/
replace charcoal retaining screens and changes surveillance 
intervals from 18 months to 24 months. Prior to the extended LCO on 
a given train, the scheduled monthly surveillance and preventive 
maintenance will be performed. This Technical Specification change 
does not involve components that are accident initiators and 
therefore will not create a new or different kind of accident than 
those previously analyzed.
    3) Involve a significant reduction in the margin of safety 
because:
    The purpose of CREFS trains A and B are to control the onsite 
dose to personnel in the Control Room following an accident that 
involves a potential radiological release. Redundant filter trains 
are utilized to ensure that a single active failure will not impact 
the ability of the system to perform its safety function. Since the 
probability of an accident occurring during the extended Technical 
Specification LCO for the inoperable train in conjunction with the 
probability that the operable CREFS train will fail is the same 
order of magnitude as for the current LCO, then the proposed 
Technical Specification change has minimal impact on the safe 
operation of the plant. The CREFS trains were both determined 
operable following their last surveillance and no events have 
occurred at the plant to indicate that they may be inoperable. 
Normal preventative maintenance and testing will be performed on the 
operable CREFS train just prior to taking the [opposite] filter 
train out of service for the modification. This action will ensure 
that the remaining subsystem is operable and ensure maximum 
reliability of the system. The change in surveillance intervals from 
18 months to 24 months will not cause a significant reduction in the 
margin of safety, because the previous five surveillances have been 
satisfactory and the equipment/components do not have a tendency to 
drift over time. Therefore, the proposed amendment will not 
significantly impact 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 
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

Dairyland Power Cooperative (DPC), Docket No. 50-409, LaCrosse 
Boiling Water Reactor (LACBWR), Vernon County, Wisconsin

    Date of amendment request: April 10, 1996
    Description of amendment request: The proposed amendment would 
update the facility Possession Only License and Technical 
Specifications to reflect the permanently shutdown and defueled 
condition of the plant.

[[Page 42279]]

    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:
    DPC proposes to modify the LACBWR Technical Specifications to 
more accurately reflect the permanently shutdown, defueled, 
possession-only status of the facility.
    Analysis of no significant hazards consideration:
    1. The proposed changes do not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes delete system requirements that are no 
longer necessary to prevent, or mitigate the consequences of, a 
credible SAFSTOR accident as described in our current SAFSTOR 
Accident Analysis.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes are either administrative in nature or were 
made based on the analysis of previously evaluated accident 
scenarios. In no other way do they change the design or operation of 
the facility and therefore do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed changes do not result in a significant reduction 
in the margin of safety.
    The changes incorporate into the proposed Technical 
Specifications the margin of safety associated with the current 
SAFSTOR accident analysis and thus don't 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: LaCrosse Public Library, 800 
Main Street, LaCrosse, Wisconsin 54601.
    Attorney for licensee: Wheeler, Van Sickle and Anderson, Suite 801, 
25 West Main Street, Madison, Wisconsin 53703-3398
    NRC Project Director: Seymour H. Weiss

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

    Date of amendment request: July 25, 1996 (NRC-96-0064)
    Description of amendment request: The proposed amendment would 
relocate or delete a number of items currently in the Administrative 
Controls Section (Section 6.0) of the technical specifications (TS). 
This submittal revises a previous submittal dated December 15, 1994 
(NRC-94-0107), to modify the proposed TS change to be consistent with 
NRC Administrative Letter 95-06, ``Relocation of Technical 
Specifications Administrative Controls Related to Quality Assurance,'' 
the Improved Standard TS (ISTS), and pending changes to the ISTS. The 
previous submittal was noticed in the Federal Register on June 6, 1995 
(60 FR 29873).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed changes are administrative in nature. None of 
the proposed changes involve a physical modification to the plant, a 
new mode of operation or a change to the UFSAR [Updated Final Safety 
Analysis Report] transient analyses. No Limiting Condition for 
Operation, ACTION statement or Surveillance Requirement is affected 
by any of the proposed changes.
    Also, these proposed changes, in themselves, do not reduce the 
level of qualification or training such that personnel requirements 
would be decreased. Therefore, this change is administrative in 
nature and does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Further, the proposed changes do not alter the design, function, or 
operation of any plant component and therefore, do not affect the 
consequences of any previously evaluated accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the proposed changes do not introduce a new mode of plant 
operation, surveillance requirement or involve a physical 
modification to the plant. The proposed changes are administrative 
in nature. The changes propose to revise, delete or relocate the 
stated administrative control provisions from the TS to the UFSAR, 
plant procedures or the QA [Quality Assurance] Program whereby, 
adequate control of information is maintained. Further, as stated 
above, the proposed changes do not alter the design, function, or 
operation of any plant components and therefore, no new accident 
scenarios are created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety because they are administrative in nature. 
None of the proposed changes involve a physical modification to the 
plant, a new mode of operation or a change to the UFSAR transient 
analyses. No Limiting Condition for Operation, ACTION statement or 
Surveillance Requirement is affected. The proposed changes do not 
involve a significant reduction in a margin of safety. Additionally, 
the proposed change does not alter the scope of equipment currently 
required to be OPERABLE or subject to surveillance testing nor does 
the proposed change affect any instrument setpoints or equipment 
safety functions. Therefore, the 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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: Mark Reinhart

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: April 29, 1996
    Description of amendment request: The proposed amendment revises 
the permissible values of the maximum and minimum pressurizer water 
levels and incorporates a graph to display these values for various 
operating conditions. The amendment also revises the Bases section of 
the Technical Specification. The Bases changes revise the acceptable 
value of the as-found tolerance for the settings of the pressurizer 
safety valves and change the value of flowrate through the pressurizer 
safety valves. The moderator temperature coefficient as described in 
the Bases Section is removed.
    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 startup accident and the rod withdrawal accident have been 
reanalyzed to justify the proposed increase in pressurizer coder 
safety value as-found tolerance. The analyses establish more 
appropriate boundaries and re-analyze the same initiators as are 
currently found in the ANO-1 Safety Analysis Report. Changing the 
as-found setpoint tolerance does not change how the pressurizer code 
safety valve operates as it will continue to be reset to 2500 psig 
plus or minus 1% prior to reactor startup.
    The acceptance criteria for these analyses are that the reactor 
coolant system (RCS)

[[Page 42280]]

pressure shall not exceed the safety limit of 2750 psig (110% of 
design pressure and that the reactor thermal power remains below 
112% Rated Power. The analyses using the proposed setpoint tolerance 
have shown that the acceptance criteria were met and that the 
consequences of the events were essentially the same as those in the 
ANO-1 SAR. Analyses were performed to determine the pressurizer 
maximum water level that would prevent the RCS from exceeding the 
safety limit of 2750 psig in the event of either a startup accident 
or a rod withdrawal accident. More appropriate pressurizer level 
requirements have been incorporated in accordance with these 
analyses.
    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 changes introduce no new mode of plant operation. 
The reanalysis of the startup accident and the rod withdrawal 
accident were performed using methodologies identical to that 
employed in the ANO-1 SAR and an improved computer code (RELAP5/
MOD2). The pressurizer code safety valve setpoint will continue to 
be reset at 2500 psig plus or minus 1% prior to reactor startup and 
will continue to function to maintain RCS pressure below the safety 
limit of 2750 psig. Analyses were performed to determine the 
pressurizer maximum water level that would prevent the RCS from 
exceeding the safety limit of 2750 psig in the event of either a 
startup accident or a rod withdrawal accident. More appropriate 
pressurizer level requirements have been incorporated in accordance 
with these analyses.
    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 safety function of the pressurizer code safety valves is not 
altered as a result of the proposed change in setpoint tolerance. 
The reanalysis of the startup accident and rod withdrawal accident 
have shown that with a plus or minus 3% setpoint tolerance, the 
pressurizer code safety valves will function to limit RCS pressure 
below the safety limit of 2750 psig. The sensitivity studies for the 
startup accident showed the acceptance criteria would still be met 
even if one pressurizer code safety valve lifted at 5% above 2500 
psig at startup conditions. Additional analyses were performed to 
determine the pressurizer maximum water level that would prevent the 
RCS from exceeding the safety limit of 2750 psig in the event of 
either a startup accident or a rod withdrawal accident.
    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 Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: June 28, 1996
    Description of amendment request: The proposed amendments would 
remove the Unit 1 and Unit 2 Technical Specification requirements to 
secure the containment equipment hatch during core alterations or fuel 
handling.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change would allow the containment equipment hatch 
door to remain open during fuel movement and core alterations. This 
door is normally closed during this time period in order to prevent 
the escape of radioactive material in the event of a fuel handling 
accident. This door is not an initiator of any accident. The 
probability of a fuel handling accident is unaffected by the 
position of the containment equipment hatch door. The current fuel 
handling analysis, which has been approved by the Staff for ANO-2 
and submitted for ANO-1, calculates maximum offsite doses to be well 
within the limits of 10 CFR Part 100. The current fuel handling 
accident analysis results in maximum offsite doses of 63.6 and 41.8 
Rem to the Thyroid and 0.902 and 0.598 Rem to the whole body (sum of 
beta and gamma) for ANO-1 and ANO-2, respectively. This analysis 
assumes the entire release from the damaged fuel is allowed to 
migrate to the site boundary unobstructed. Therefore, allowing the 
equipment hatch doors to remain open results in no change in 
consequences. Also, the calculated doses during a fuel handling 
accident would be considerably larger than the actual doses since 
the calculation does not incorporate the closing of the equipment 
hatch door following evacuation of containment. The proposed change 
would significantly reduce the dose to workers in the containment in 
the event of a fuel handling accident by expediting the containment 
evacuation process. Therefore, this change does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change would 
not alter the design, configuration, or method of operation of the 
plant beyond the standard functional capabilities of the equipment. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed change does not have the potential for an increased 
dose at the site boundary due to a fuel handling accident. The 
margin of safety as defined by 10 CFR Part 100 has not been 
significantly reduced. Closing the equipment hatch door following an 
evacuation of containment further reduces the offsite doses in the 
event of a fuel handling accident and provides additional margin to 
the calculated offsite doses. 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

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

    Date of amendment request: July 12, 1996
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Sections 6.2.2.h and 6.2.2.i. To 
provide adequate shift coverage without routine heavy use of overtime, 
TS Section 6.2.2.h specifies an objective to have operating personnel 
work ``a normal 8-hour day, 40-hour week'' while the facility is 
operating. The proposed amendment would change the objective to ``an 8 
to 12 hour day, nominal 40-hour week.''
    TS Section 6.2.2.i currently states, ``The General Supervisor 
Operations, Supervisor Operations, Station Shift Supervisor Nuclear, 
and Assistant Station Shift Supervisor Nuclear shall hold senior 
reactor operator licenses.'' The proposed amendment would change this 
section to state, ``The

[[Page 42281]]

Manager Operations, Station Shift Supervisor Nuclear and Assistant 
Station Shift Supervisor Nuclear shall hold senior reactor operator 
licenses.'' This change is based upon a reorganization that eliminates 
the positions of General Supervisor Operations and Supervisor 
Operations from the Unit 1 Operations management structure. The 
responsibilities of these positions will be assumed by the Manager 
Operations or delegated to off-shift Senior Reactor Operators. Thus, 
Senior Reactor Operators will report directly to the Manager 
Operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40-hour week,'' provides enhanced continuity for 
normal plant operations. There has been no noticeable increase in 
safety related problems during the trial period [The facility has 
been implementing 12-hour operator shifts for over 1 year on a trial 
basis]. Overtime remains controlled by site administrative 
procedures in accordance with the NRC Policy Statement of working 
hours (Generic Letter 82-12). The probability for operating 
personnel error due to (1) incomplete or insufficient turnover or 
(2) interruption of in-plant maintenance and testing is reduced. No 
physical plant modifications are involved, and none of the 
precursors of previously evaluated accidents are affected. 
Therefore, this change will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The assimilation of the responsibilities of the previous 
positions of General Supervisor Operations and Supervisor Operations 
into the position of Manager Operations and to off-shift Senior 
Reactor Operators reflects a restructuring of the operations 
department, and is essentially a reduction in layers of management. 
This proposed change does not involve any physical modification to 
the plant, and does not affect any precursor of a previously 
evaluated accident. Therefore, this change will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Establishing operating personnel hours at ``an 8 to 12-hour day, 
nominal 40-hour week'' provides increased flexibility in scheduling 
and does not adversely affect their performance. Overtime remains 
controlled by site administrative procedures in accordance with the 
NRC Policy Statement on working hours (Generic Letter 82-12). No 
physical modification of the plant is involved. As such, the change 
does not introduce any new failure modes or conditions that may 
create a new or different accident. Therefore, operation in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The responsibilities of the previous positions of General 
Supervisor Operations and Supervisor Operations will be assimilated 
into the positions of the Manager Operations and the off-shift 
Senior Reactor Operators. There is no physical plant modification. 
The change does not introduce any new failure modes or conditions 
that may create a new or different accident. Therefore, the change 
does not in itself create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    Establishing operating personnel hours at ``an 8 to 12-hour day, 
nominal 40-hour week,'' provides increased flexibility in scheduling 
and does not adversely affect their performance. This change also 
decreases the risk of miscommunication between shifts by reducing 
the number of turnovers per day and increases operations and 
maintenance efficiency by promoting continuity in ongoing plant 
activities. Overtime remains controlled by site administrative 
procedures in accordance with the NRC Policy Statement on working 
hours (Generic Letter 82-12) and is consistent with the Improved 
Standard Technical Specifications. The proposed change involves no 
physical modification of the plant, or alterations to any accident 
or transient analysis [...], and the changes are administrative in 
nature. Therefore, the change does not involve any significant 
reduction in a margin of safety.
    The assimilation of the responsibilities of the positions of 
General Supervisor Operations and Supervisor Operations, into the 
positions of the Manager Operations and the off-shift Senior Reactor 
Operators, effectively reduces layers of management. The proposed 
change is consistent with Standard Review Plan (SRP) 13.1.2-13.1.3. 
This administrative transformation of the operations department 
management structure involves no physical modification of the plant 
or alterations to any accident or transient analysis. Therefore, 
this change in itself does not involve any 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
    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: Jocelyn A. Mitchell, Acting Director

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

    Date of amendment request: July 12, 1996
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 6.2.2.i. To provide 
adequate shift coverage without routine heavy use of overtime, TS 
Section 6.2.2.i specifies an objective to have operating personnel work 
``a normal 8-hour day, 40-hour week'' while the facility is operating. 
The proposed amendment would change the objective to ``an 8 to 12 hour 
day, nominal 40-hour week.''
    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.
    Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40-hour week,'' allows normal plant operations to 
be managed more effectively and with enhanced continuity. There has 
been no noticeable increase in safety related problems during the 
trial period [The facility has been implementing 12-hour operator 
shifts for over 1 year on a trial basis]. Overtime remains 
controlled by site administrative procedures in accordance with the 
NRC Policy Statement on working hours (Generic Letter 82-12). The 
probability for operating personnel error due to (1) incomplete or 
insufficient turnover or (2) interruption of in-plant maintenance 
and testing is reduced. No physical plant modifications are 
involved, and none of the precursors of previously evaluated 
accidents are affected. Therefore, this change will not involve a 
significant increase in the probability or 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.
    Establishing operating personnel hours at, ``an 8 to 12-hour 
day, nominal 40-hour week,'' improves the quality of life for 
operating personnel and does not adversely affect their performance. 
Overtime remains controlled by site administrative procedures in 
accordance with the NRC Policy Statement on working hours (Generic 
Letter 82-12). No physical modification of the plant is

[[Page 42282]]

involved. As such, the change does not introduce any new failure 
modes or conditions that may create a new or different accident. 
Therefore, operation in accordance with the proposed amendment 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.
    Establishing operating personnel hours at, ``an 8 to 12-hour 
day, nominal 40-hour week,'' improves the quality of life for 
operating personnel and does not adversely affect their performance. 
This change also decreases the risk of miscommunication between 
shifts and increases operations and maintenance efficiency by 
promoting continuity in ongoing plant activities. Overtime remains 
controlled by site administrative procedures in accordance with the 
NRC Policy Statement on working hours (Generic Letter 82-12) and is 
consistent with the Improved Standard Technical Specifications. The 
proposed change involves no physical modification of the plant, or 
alterations to any accident or transient analysis [...], and the 
changes are administrative in nature. Therefore, the change does not 
involve any 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: Jocelyn A. Mitchell, Acting Director

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: February 2, 1996
    Description of amendment request: This request would change 
Technical Specification (TS) 3.6.1.2 for each unit to permit primary 
containment leakage testing of the main steam isolation valves (MSIVs) 
at either 22.5 psig or 45 psig according to the type of test to be 
conducted. Currently the TS only specifies 22.5 psig for the MSIVs' 
test pressure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the allowable test pressure for MSIV leak 
testing was reviewed from two perspectives. First is the potential 
for the change in testing pressure, and test methodology, to impact 
testing results. The second perspective is the potential for a 
failure of the testing configuration to result in undesirable 
consequences.
    Under the proposed change, an increased test pressure of 45.0 
psig (Pa) in the accident direction will be used to perform 
Technical Specification required MSIV leak testing. However, the 
acceptance criteria for testing is maintained consistent with 
current Technical Specifications. Therefore, the proposed change to 
allow a test pressure of Pa will not affect the validity of 
leak test results. The existing Technical Specification required 
leak integrity of the MSIVs will be maintained under the proposed 
test methodology and thus the ability of the MSIVs to act as a 
containment isolation valves is not affected.
    The proposed test pressure of Pa will be applied in the 
accident direction, and will result in a back pressure being applied 
to the Main Steam Line (MSL) Plugs. The potential for MSL Plug 
ejection has been reviewed and adequate precautions have been taken 
to ensure that fuel damage would not result from [local leak rate 
test] LLRT induced MSL Plug ejection. The MSL Plugs are installed 
using a restraint ring which prevents inadvertent ejection. 
[Pennsylvania Power and Light Company] PP&L procedures require that 
the restraint ring be installed as a prerequisite for LLRT testing 
of the MSIVs at Pa. However, in the unlikely event that the MSL 
Plug and restraint ring were installed improperly and then subjected 
to back pressurization at Pa, ejection could occur. If this 
event did occur, the MSL Plug could hit the fuel which is an 
accident bounded by the fuel assembly handling accident analysis 
addressed in [Final Safety Analysis Report] FSAR Section 15.7.4. The 
MSL Plugs, MSL Plug Restraint Ring, and MSL Plug Insert and Remove 
Tool meet the requirements of NUREG 0612 and PP&L's Heavy Loads 
Program.
    Therefore, the proposal to allow an alternative test pressure, 
Pa, does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    All components within the test volume have been evaluated for 
structural integrity under the proposed test pressures. In addition, 
pressurization of the Main Steam Line Plugs during testing will be 
below the evaluated pressure. The acceptance criteria for the test 
will be maintained, thus verification of the leak integrity of the 
MSIVs will not be impacted. Therefore, the proposed change to allow 
for an alternative test pressure of (Pa) does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    The proposed change does not affect the acceptance criteria for 
the MSIV LLRT. As a result, testing at Pa in the accident 
direction will provide an equivalent test to that which is performed 
at Pa. No change in the leak integrity of the MSIVs is 
anticipated as a result of performing the testing at the alternative 
pressure. The potential for MSL Plug ejection during MSIV LLRT at 
Pa has been evaluated and found to be bounded by existing 
accident analysis. Therefore the proposed change to allow an 
alternative test pressure, Pa, 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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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: July 12, 1996
    Description of amendment request: The proposed amendment would 
revise the Indian Point 3 (IP3) Technical Specifications (TSs) by 
changing the surveillance frequency requirements in Table 4.1-1, 
``Minimum Frequencies for Checks, Calibrations, and Tests of Instrument 
Channels'' to accommodate a 24-month operating cycle.
    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 proposed changes do not involve a significant increase in 
the probability or consequence of any accident previously evaluated. 
The proposed changes are being made to extend surveillance 
frequencies from 18 months to 24 months for:

[[Page 42283]]

    Vapor Containment High Radiation Monitors
    Reactor Coolant System Subcooling Margin Monitor (SMM),
    Overpressure Protection System (OPS), and
    Reactor Vessel Level Indication System (RVLIS).
    These proposed changes are being made using the guidance 
provided by Generic Letter 91-04 to accommodate a 24-month fuel 
cycle. The containment radiation monitors, SMM, and RVLIS are used 
to provide operator information during post-accident conditions and 
have no effect on event initiators associated with previously 
analyzed accidents. The OPS is used only when the plant is shutdown, 
with RCS [reactor coolant system] temperature below a low 
temperature limit, and the RCS is not vented. The function of the 
OPS is to protect the RCS from Low Temperature Overpressurization 
(LTOP) transients and has no effect on accident initiators. No 
credit is taken in the IP3 safety analyses for accident mitigation 
effects that might result from use of these instrument channels. 
Updated calculations and evaluations to assess the proposed increase 
in the surveillance intervals demonstrate that the effectiveness of 
these instrument channels in fulfilling their respective functions 
is not reduced. The containment high radiation monitors are used for 
post accident monitoring purposes to provide operators with an 
indication of adverse conditions in containment based on releases of 
radioactivity from the RCS to the containment atmosphere. These 
monitors provide no signals to plant control systems or automatic 
safety systems used for accident mitigation and have no role as an 
accident initiator.
    Use of the subcooling margin monitor and core exit thermocouples 
by plant operators is specified in the Indian Point 3 Emergency 
Operating Procedures (EOPs) to assess post accident cooling 
conditions in the RCS. Changes to the EOPs will be made to reflect 
the results of the updated loop accuracy calculations for this 
instrumentation. These changes will ensure that safety analysis 
input assumptions associated with subcooling margin, for small break 
LOCA [loss-of-coolant accident], steam generator tube rupture, and 
steamline break, remain valid, and that the response strategies 
outlined in the Westinghouse Owners Group Emergency Response 
Guidelines are maintained. Core exit thermocouple readings are not 
used for input to plant safety analyses.
    The OPS provides a protective function to prevent RCS pressure 
limits from being exceeded while the plant is shutdown and the RCS 
is being maintained at a low temperature and not vented. Failure of 
the OPS is not assumed to be an accident initiator in the plant 
safety analyses.
    The change to the RVLIS calibration interval does not affect 
design or operation of plant systems and will not affect the 
probability of accidents. Revised loop accuracy calculations have 
demonstrated that operator actions for responding to postulated 
accidents using RVLIS in conjunction with the Indian Point 3 EOPs 
will remain consistent with the accuracy requirements RVLIS. The 
consequences of a previously evaluated accident will not be 
affected.
    Equipment and system design requirements and safety analysis 
acceptance criteria continue to be met with the proposed new 
surveillance intervals. Based on the above information it is 
concluded that 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 changes to extend the surveillance frequencies for 
the above listed instrument channel do not create the possibility of 
a new or different kind of accident from any previously evaluated. 
The increased surveillance frequencies were evaluated based on past 
equipment performance and do not require any plant hardware changes 
or changes in system operation. There are no new failure modes 
introduced as a result of extending these surveillance intervals, 
which could lead to the creation of new or different kinds of 
accident.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed changes do not involve a significant reduction in a 
margin of safety. [A decreased] surveillance frequency for the 
Containment High Radiation Monitor, SMM, OPS, and RVLIS does not 
adversely affect the performance of safety-related systems, 
equipment, or instruments and does not result in increased severity 
of accidents evaluated. The radiation monitor, SMM, and RVLIS are 
not used to support margins of safety identified in the Technical 
Specifications. OPS provides an equipment protection function to 
prevent inadvertent overpressurization of the RCS at shutdown 
conditions. The Low Temperature Overpressurization (LTOP) curve in 
the Technical Specifications represents material stress limits based 
on fracture toughness requirements for ferritic steel. Analysis of 
the proposed change to the OPS surveillance frequency verified 
sufficient margin to the LTOP curve and therefore does not involve a 
significant reduction in margin to the material stress limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Jocelyn A. Mitchell, Acting Director

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: July 12, 1996
    Description of amendment request: The proposed amendment would 
change the Indian Point 3 (IP3) Technical Specifications (TS) relating 
to minimum reactor coolant system (RCS) flow and maximum RCS average 
temperature to make these parameters consistent with an assumption of 
100% helium release from the boron coating of the integral fuel 
burnable absorber (IFBA) rods.
    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 any accident 
previously evaluated?
    The proposed changes to the RCS minimum flow and maximum 
Tavg requirements will not increase the probability or 
consequences of an accident previously evaluated. Reference 2 [SECL-
96-046, ``IFBA Helium Release Evaluation for Cycle 9 Restart,'' 
Westinghouse Electric Corporation, dated July 8, 1996] states that, 
for the remainder of Cycle 9, all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the Technical Specification Bases is not reduced in any 
of the licensing basis accident analyses for the assumption of a 
100% helium release from the IFBA rods. Reference 3 [Westinghouse 
letter, ``Technical Specification Value for T-Average,'' INT-96-557, 
dated July 3, 1996] states that a reduction of maximum allowable 
indicated Tavg from 578.3 deg.F to 571.5 deg.F specifications 
consistent with the more limiting containment integrity analyses. 
The associated plant and technical specification changes do not 
affect any of the mechanisms postulated in the FSAR [Final Safety 
Analysis Report] to cause licensing basis events. Therefore, the 
probability of an accident previously evaluated has not increased. 
Because design limitations continue to be met, and the integrity of 
the RCS pressure boundary is not challenged, the assumptions 
employed in the calculation of the offsite radiological doses remain 
valid. Therefore, the consequences of an accident previously 
evaluated will not be increased.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any previously 
evaluated?
    The proposed changes to the RCS minimum flow and maximum 
Tavg requirements do not create the possibility of a new or 
different kind of accident from any previously evaluated. Reference 
2 states that, for the remainder of Cycle 9, all pertinent

[[Page 42284]]

licensing basis acceptance criteria have been met, and the margin of 
safety as defined in the Technical Specification Bases is not 
reduced in any of the licensing basis accident analyses for the 
assumption of a 100% helium release from the IFBA. Reference 3 
provides clarifications of the assumptions made in the design basis 
and restricts DNB temperature limits to be consistent with non-DNB 
analyses. The associated plant and technical specification changes 
do not change the plant configuration in a way which introduces a 
new potential hazard to the plant (i.e., no new failure mode has 
been created). Therefore, an accident which is different than any 
previously evaluated will not be created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed changes to the RCS minimum flow and maximum 
Tavg requirements do not involve a significant reduction in a 
margin of safety. Reference 2 demonstrates that, for the remainder 
of Cycle 9, all pertinent licensing basis acceptance criteria have 
been met, and the margin of safety as defined in the Technical 
Specification Bases is not reduced in any of the licensing basis 
accident analyses for the assumption of a 100% helium release from 
the IFBA. Reference 3 maintains the margin of safety by restricting 
a DNB limit to bound other analyses. Since References 2 and 3 
demonstrate that all applicable acceptance criteria continue to be 
met, the subject operating conditions 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: 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: Jocelyn A. Mitchell, Acting

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

    Date of amendment request: May 3, 1996 (TS 352)
    Description of amendment request: The proposed amendment requests 
administrative changes to the Browns Ferry Nuclear Plant (BFN) Units 1, 
2, and 3 technical specifications. The proposed amendment consists of 
three parts, designated by the licensee as A, B, and C. Part A deletes 
technical specification requirements associated with BFN Unit 2 
Amendment 219, issued November 12, 1993, to permit modification of 
reactor vessel water level instrumentation requested by NRC Bulletin 
93-03. Part B deletes technical specification requirements associated 
with Amendment 228, issued on December 7, 1994, which provided a 
temporary change to permit upgrade of electrical equipment. The 
modifications associated with Parts A and C are complete. Part C 
provides other administrative changes to clarify requirements and to 
implement rule 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 proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Part A: The proposed Technical Specification change to remove 
the temporary revisions, which were in place to modify the reactor 
vessel water level instrumentation requested by NRC Bulletin 93-03, 
is administrative. The temporary limiting condition for the minimum 
number of trip systems operable will no longer be accurate and the 
minimum number operable per trip system will be the same as they 
were prior to November 12, 1993. Therefore, the proposed changes 
will not significantly increase the consequences of an accident 
previously evaluated.
    Part B: The proposed Technical Specification change to remove 
the temporary revisions, which were in place to replace the 250 volt 
shutdown board batteries is administrative. The LCO to extend the 
allowed outage time (AOT) from a five-day to a 45-day AOT will no 
longer be accurate and the five day AOT will be the same as it was 
prior to Unit 2, Cycle 7. Therefore, the proposed changes will not 
significantly increase the consequences of an accident previously 
evaluated.
    Part C: The proposed Technical Specifications change revises 
items 1 through 5 above (Section I, Description of the Proposed 
Change, Part C), and is administrative. TVA has evaluated the 
proposed technical specification changes and has determined that the 
proposed changes are administrative in nature. Further, it provides 
a revision based on an NRC Code of Federal Regulations rule change. 
Also, the proposed changes provide correction of administrative 
errors from previous technical specifications. For example, the Main 
Steamline High Radiation remarks in Table 3.2.A, 1.b., should have 
been deleted from the TS as part of TS-322. It also clarifies some 
requirements to ensure consistent application throughout the 
specifications. These changes do not affect any of the design basis 
accidents. They do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Part A: The proposed Technical Specification change to remove 
the temporary revisions, which were in place to modify the reactor 
vessel water level instrumentation requested by NRC Bulletin 93-03, 
is administrative. The temporary limiting condition for the minimum 
number of trip systems operable will no longer be accurate and the 
minimum number operable per trip system will be the same as they 
were prior to November 12, 1993. No modifications to any plant 
equipment are involved. There are no effects on system interactions 
made by these changes. They do not create the possibility of a new 
or different kind of accident from an accident previously evaluated.
    Part B: The proposed Technical Specification change to remove 
the temporary revisions, which were in place to replace the 250 volt 
shutdown board batteries is administrative. The LCO to extend the 
allowed outage time (AOT) from a five day to a 45-day AOT will no 
longer be accurate and the five day AOT will be the same as it was 
prior to Unit 2, Cycle 7. No modifications to any plant equipment 
are involved. There are no effects on system interactions made by 
these changes. They do not create the possibility of a new or 
different kind of accident from an accident previously evaluated.
    Part C: The proposed Technical Specifications change revises 
items 1 through 5 above (Section I, Description of the Proposed 
Change, Part C), and is administrative. TVA has evaluated the 
proposed changes and has determined that they are administrative in 
nature. Further, it provides revisions based on an NRC Code of 
Federal Regulations rule change. It also provides correction of 
administrative errors in previous technical specification changes. 
For example, the Main Steamline High Radiation remarks in Table 
3.2.A, 1.b., should have been deleted from the TS as part of TS-322. 
It also clarifies some requirements to ensure consistent application 
throughout the specifications. These changes do not affect any of 
the design basis accidents. No modifications to any plant equipment 
are involved. There are no effects on system interactions made by 
these changes. They do not create the possibility of a new or 
different kind of accident from an accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change is administrative in nature for Parts A, B, 
and C. The proposed change includes the deletion of temporary 
changes as a result of modifications to systems and clarification of 
some requirements to ensure consistent application throughout the 
specifications. Further, the proposed change corrects errors in 
previous TS submittals. No safety margins are affected by these 
changes.
    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

[[Page 42285]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street,Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon
    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 l0H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: June 21, 1996 (TS 377)
    Description of amendment request: The proposed amendment provides a 
new minimum critical power ratio safety limit to replace the current 
non-conservative value. The amendment also updates the technical 
specification bases to clarify the usage of the residual heat removal 
supplemental spent fuel pool cooling mode.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change in the Safety Limit Minimum Critical Power 
Ratio (SLMCPR) does not increase the frequency of the precursors to 
design basis events or operational transients analyzed in the Browns 
Ferry Final Safety Analysis Report. Therefore, the probability of an 
accident previously evaluated is not significantly increased.
    The proposed change in the SLMCPR ensures that 99.9 percent of 
the fuel rods in the core are expected to avoid boiling transition 
during the most limiting anticipated operational occurrence, which 
is the design and licensing basis for the analysis of accidents and 
transients described in the Browns Ferry Updated Final Safety 
Analysis Report (UFSAR). It does not change the nuclear safety 
characteristics of any safety system or containment system. 
Therefore, the consequences of an accident, operator error, or 
malfunction of equipment important to safety previously evaluated in 
the UFSAR has not been increased.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to the Technical Specification requirements 
for the safety limit minimum critical power ratio does not involve a 
modification to plant equipment. No new failure modes are 
introduced. There is no effect on the function of any plant system 
and no new system interactions are introduced by this change. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will ensure that during any anticipated 
operational transient, at least 99.9% of the fuel rods would be 
expected to avoid boiling transition which is consistent with the 
licensing basis. Since the margin [of] safety is being increased 
with this change, the proposed amendment does not involve a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, 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: July 18, 1996
    Description of amendment request: The amendment adopts ASTM D-3803-
1989 as the laboratory testing standard for charcoal samples from the 
charcoal adsorbers in the auxiliary/fuel building emergency exhaust 
system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The requested change to the charcoal sample surveillance 
acceptance criteria for the fuel building and auxiliary building 
emergency exhaust system will not affect the method of operation of 
the system. The testing of the charcoal filter samples will continue 
to be performed in accordance with NRC-accepted methods and 
acceptance criteria, and the new test protocol will still ensure 
filter efficiency is maintained equal to or greater than 90%. There 
are no changes to the emergency exhaust system and it will continue 
to function in a manner consistent with the safety analysis 
assumptions and the plant design basis. There will be no degradation 
in the performance of or an increase in the number of challenges to 
equipment assumed to function during an accident. Therefore, the 
proposed changes 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.
    The changes to the surveillance requirements are being made to 
adopt current NRC-accepted methods of testing charcoal samples. 
These changes will not affect the method of operation of the 
applicable systems and the laboratory testing will continue to 
demonstrate the required adsorber performance after a design-basis 
LOCA [loss-of-coolant accident] or fuel handling accident. No new or 
different kind of accident from any previously evaluated will be 
created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The new charcoal adsorber sample laboratory testing protocol is 
more stringent than the current testing practice and meets current 
NRC-approved test methods. The new testing criteria will continue to 
demonstrate the required adsorber performance after a design-basis 
LOCA or fuel handling accident and will not affect the filter system 
performance. Therefore, this change will not reduce the margin of 
safety of the emergency exhaust system filter operation.
    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, NW., Washington, DC 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: July 18, 1996
    Description of amendment request: The proposed amendment would 
revise

[[Page 42286]]

Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 3.8, 
``Refueling Operations,'' and its associated Basis, by allowing the 
containment personnel air lock doors to remain open during refueling 
operations as long as at least one door is capable of being closed in 
30 minutes or less.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to determine that no significant hazards 
exist. The proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Maintaining the doors of the personnel air lock open during 
REFUELING OPERATIONS does not adversely affect the probability or 
consequences of accidents previously evaluated. The only applicable 
accident is a fuel handling accident described in [Updated Safety 
Analysis Report] USAR Section 14.2.1. The fuel handling accident 
evaluated in the USAR Section 14.2.1 assumes the accident to be in 
the spent fuel pool in the Auxiliary Building. The accident assumes 
a sudden release of the gaseous fission products held in the voids 
between the pellets and cladding of all of the rods in the highest 
rated fuel assembly at 100 hours following reactor shutdown. The 
accident activity is assumed to discharge from the spent fuel pool 
directly to the atmosphere at ground level. No credit is taken for 
existing building structures, ventilation, or filtration systems. A 
fuel handling accident in containment is bounded by this evaluation. 
Furthermore, any release from a fuel handling accident in 
containment can still be terminated by closing one of the personnel 
air lock doors following containment evacuation.
    The containment personnel air lock doors are components integral 
to the containment structure. They are not accident initiators. 
Therefore, the proposed amendment does not increase the probability 
of any previously evaluated accident.
    The control room operator immersion and inhalation doses were 
reviewed as part of the updated Control Habitability Evaluation 
Report. The report states that thyroid and whole body doses received 
by control room operators in each of the other design basis 
accidents discussed in KNPP USAR Section 14.2 are less than the 
[loss of coolant accident] LOCA dose. This amendment does not change 
the results of the Control Room Habitability Evaluation Report, 
since the fuel handling accident evaluated in KNPP USAR Section 
14.2.1 assumes a release directly to the atmosphere. This change 
does not significantly increase the consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The accident evaluated in USAR section 14.2.1 bounds a fuel 
handling accident in containment with the personnel air lock doors 
open. The fuel handling accident evaluated in USAR section 14.2.1 
assumes activity is discharged directly to the atmosphere at ground 
level. Since no credit is taken for building structures, ventilation 
systems or filtration systems, the position of the doors does not 
affect the analysis of record. Furthermore, one of the air lock 
doors can still be closed following containment evacuation to 
terminate the release.
    The containment personnel air lock doors are components integral 
to the containment structure. They are not accident initiators. The 
proposed amendment does not create the possibility of any new or 
different kind of accident [from any accident] previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    Maintaining the containment personnel air lock doors open during 
REFUELING OPERATIONS does not involve a significant reduction in the 
margin of safety. A fuel handling accident in containment is bounded 
by a fuel handling accident in the spent fuel pool. The spent fuel 
pool fuel handling accident is assumed to have a sudden release of 
the gaseous fission products held in the voids between the pellets 
and cladding of all of the rods in the highest rated fuel assembly, 
100 hours following reactor shutdown. The accident activity leaving 
the spent fuel pool is assumed to discharge directly to the 
atmosphere at ground level. No credit is taken for existing building 
structures, ventilation, and filtration systems. Therefore, there is 
no reduction in the current margin of safety. Furthermore, the 
release caused by a fuel handling accident in containment can be 
terminated by closing one of the personnel air lock doors following 
containment evacuation.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

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

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

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: July 12, 1996
    Brief description of amendment request: The amendment would change 
Technical Specification 3.3.2.1, ``Engineered Safety Feature Actuation 
System Instrumentation,'' to reflect a revised setpoint for the 
interlock designated P-12.
    Date of publication of individual notice in Federal Register: July 
23, 1996 (61 FR 38229)
    Expiration date of individual notice: August 22, 1996
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

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

[[Page 42287]]

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.

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

    Date of application for amendment: January 29, 1996, as 
supplemented June 17, 1996.
    Brief description of amendment: The amendment revises the technical 
specifications (TS) table 4.1-3, item 4 to change the frequency of main 
steam safety valve (MSSV) testing to that specified in NUREG-1431, the 
improved ``Standard Technical Specifications, Westinghouse Plants'' and 
adds the MSSV test acceptance requirements.
    Date of issuance: August 1, 1996
    Effective date: August 1, 1996
    Amendment No.: 171
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7545). The June 17, 1996, submittal provided supplemental 
information that was not outside the scope of the February 28, 1996, 
notice. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 1, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

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

    Date of application for amendment: March 20, 1996
    Brief description of amendment: To relocate Technical Specification 
3.3.3.2, Movable Incore Detectors, to plant procedures.
    Date of issuance: July 24, 1996
    Effective date: July 24, 1996
    Amendment No.: 65
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18164) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 24, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: November 22, 1995
    Brief description of amendments: The amendments replace the title-
specific designation of members representing specific functional areas 
on the Plant Operating Review Committee (PORC) for the Haddam Neck 
Plant and Millstone Units 1, 2, and 3 with a functional area-specific 
designation that stipulates membership qualification and experience 
requirements. The amendments also clarify the composition of the Site 
Operations Review Committee (SORC) at Millstone.
    Date of issuance: July 16, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 190, 95, 200, 130
    Facility Operating License Nos. DPR-61, DPR-21, DPR-65, AND NPF-49: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7549) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 16, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and 
the Learning Resources Center, Three Rivers Community-Technical 
College, 574 New London Turnpike, Norwich, Connecticut 06360, and 
Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, 
Connecticut 06385, for Millstone 1, 2, and 3.

Duke Power Company, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1, York County, South Carolina

    Date of application for amendment: January 26, 1996, as 
supplemented May 6, May 20, and June 5, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications to permit a one-time operation of the containment purge 
ventilation system during Mode 3 and 4 after the steam generator 
replacement outage.
    Date of issuance: July 30, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.: 150
    Facility Operating License No. NPF-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18165) The supplemental submittals provided clarifying information that 
did not change the scope of the January 26, 1996, application for 
amendment nor the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 30, 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. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: March 4, 1996
    Brief description of amendments: The amendments delete Flow 
Monitoring System from Technical Specification 3.4.6.1 and associated 
surveillance requirements.
    Date of issuance: July 29, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 168 and 150
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18166) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 29, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

[[Page 42288]]

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

    Date of application for amendments: March 4, 1996
    Brief description of amendments: The amendments consist of changes 
to the Final Safety Analysis Report for McGuire Units 1 and 2 to delete 
the seismic qualification requirement for the Containment Atmosphere 
Particulate Radiation Monitors.
    Date of issuance: July 30, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 169 and 151
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Final Safety Analysis Report.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20845) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 30, 1996, and an 
Environmental Assessment dated July 22, 1996. No significant hazards 
consideration comments received: No.
    Local Public Document Room location:  Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: May 20, 1996
    Brief description of amendment: The amendment revised the Facility 
Operating License and Appendix C to the license to reflect the name 
change from Gulf States Utilities Company to Entergy Gulf States, Inc.
    Date of issuance: July 30, 1996
    Effective date: July 30, 1996
    Amendment No.: 88
    Facility Operating License No. NPF-47: The amendment revised the 
operating license and Appendix C to the license.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31183) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 30, 1996. No significant 
hazards consideration comments received. No
    Local Public Document Room location:  Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

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

    Date of application for amendment: November 20, 1995, as 
supplemented by letter dated December 15, 1995
    Brief description of amendment: The amendment revised and deleted 
surveillance requirements, notes, and action statements involved with 
the requirements for the drywell leak rate testing, and the air lock 
leakage and interlock testing in Subsections 3.6.5.1 (Drywell), 3.6.5.2 
(Drywell Air Lock), and 3.6.5.3 (Drywell Isolation Valves) of the 
technical specifications.
    Date of issuance: August 1, 1996
    Effective date: August 1, 1996
    Amendment No: 126
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25704) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 1, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: March 21, 1996 as supplemented 
May 13, 1996.
    Brief description of amendments: Relocate requirements for 
Radiological Effluent Controls from Technical Specifications (TS) to 
the Offsite Dose Calculation Manual or the Process Control Program. New 
programmatic controls for radioactive effluent and radiological 
environmental controls will be incorporated into the TS. Also, 
requirements for Gas Decay tanks and Explosive Gas Mixture will be 
placed in a different area of the TS.
    Date of issuance: July 31, 1996
    Effective date: July 31, 1996
    Amendment Nos.: 188 and 182Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1966 (61 FR 
31180) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 31, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: May 28, 1996
    Brief description of amendments: Amendment changes Technical 
Specification 6.2.2.i, ``Administrative Controls,'' regarding 
Operations Manager qualifications.
    Date of issuance: July 22, 1996
    Effective date: July 22, 1996
    Amendment Nos.: 187 and 181Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31181) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 22, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

GPU Nuclear Corporation and Saxton Nuclear Experimental (SNEC) 
Corporation, Docket No. 50-146, Saxton Nuclear Experimental 
Facility (SNEF)

    Date of application for amendment: February 2, 1996, as 
supplemented on February 28, April 24, and May 24, 1996.
    Brief description of amendment: The proposed amendment would (1) 
increase the scope of work permitted at SNEF to include asbestos 
removal, removal of defunct plant electrical services, and installation 
of decommissioning support facilities and systems; (2) eliminate areas 
within the containment vessel requiring administrative access controls; 
and (3) revise the facility layout diagram to allow the exclusion area 
to consist of, at a minimum, the containment vessel and, at a maximum, 
to extend to the SNEF outer security fence and to include on the 
diagram the footprint of the proposed decommissioning support 
facilities.
    Date of issuance: July 23, 1996
    Effective date: July 23, 1996
    Amendment No.: 14
    Amended Facility License No. DPR-4: Amendment changed the Technical 
Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31182).

[[Page 42289]]

The Commission's related evaluation of the amendment is contained in a 
safety evaluation dated July 23, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678

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

    Date of amendment request: February 1, 1996
    Brief description of amendment: The amendment revised Technical 
Specifications to allow an increase in the initial nominal Uranium-235 
enrichment limit for fuel assemblies which may be stored in the spent 
fuel pool.
    Date of issuance: July 30, 1996
    Effective date: July 30, 1996
    Amendment No.: 174
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10396) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 30, 1996 . No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendments: May 9, 1996
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Nuclear 
Power Plant (DCPP), Unit Nos. 1 and 2 by revising Technical 
Specifications (TS) 3/4.3.2, ``Engineered Safety Features Actuation 
System Instrumentation,'' and 3/4.6.2, ``Containment Spray System.'' 
The changes clarified the description of the initiation signal required 
for operation of the containment spray system at DCPP and correctly 
incorporated changes made in previous license amendments. All of the 
changes are administrative in nature.
    Date of issuance: August 1, 1996
    Effective date: August 1, 1996
    Amendment Nos.: Unit 1 - 114; Unit 2 - 112
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31184) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 1, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: June 3, 1996, as superseded by 
application dated June 25, 1996.
    Brief description of amendments: These amendments revise Improved 
Technical Specification (TS) 3.3.11, ``Post Accident Monitoring 
Instrumentation (PAMI),'' and Improved TS 5.5.2.13, ``Diesel Fuel Oil 
Testing Program.'' Specifically, the number of instruments required to 
measure reactor coolant inlet temperature (TCold), and reactor 
coolant outlet temperature (THot), will be revised from two per 
loop to two (with one cold leg indication and one hot leg indication 
per steam generator). These changes to the Improved TS reinstate 
provisions of the current San Onofre Nuclear Generating Station 
(SONGS), Unit Nos. 2 and 3 TS revised as part of NRC Amendment Nos. 127 
and 116 for SONGS Units 2 and 3 (referred to as the Improved TS).
    Date of issuance: August 1, 1996
    Effective date: August 1, 1996, to be implemented by August 9, 
1996.
    Amendment Nos.: Unit 2 - 130; Unit 3 - 119
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1996 (61 FR 
34452) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 1, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

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

    Date of application for amendments: July 26, 1995, as supplemented 
April 25, 1996. The April 25, 1996, letter provided clarifying 
information that did not change the scope of the July 26, 1995, 
application and initial proposed no significant hazards consideration 
determination.
    Brief description of amendments: The amendments clarify the 
Technical Specifications to allow switching of charging and low-head 
safety injection pumps during unit shutdown conditions. These 
amendments also allow additional methods of rendering these same pumps 
incapable of injecting into the reactor coolant system when required 
for low-temperature conditions.
    Date of issuance: July 24, 1996
    Effective date: July 24, 1996
    Amendment Nos.: 202 and 183
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45190) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 24, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendment: May 8, 1996
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification (TS) 5.3, ``Reactor,'' and 
TS 5.4, ``Fuel Storage,'' by removing the enrichment limit for reload 
fuel and imposing fuel storage restrictions on the spent fuel storage 
racks and the new fuel storage racks. The revised TS are structured 
consistent with the Westinghouse Standard Technical Specifications and 
the fuel storage restrictions are based on the criticality analyses 
used to support Amendment No. 92 dated March 7, 1991.
    Date of issuance: July 23, 1996
    Effective date: July 23, 1996
    Amendment No.: 124
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31185) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 23, 1996. No significant 
hazards consideration comments received: No.

[[Page 42290]]

    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

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

    Date of amendment request: May 1, 1995
    Brief description of amendment: This amendment revises TS Section 
6.0, throughout, to reflect an organization change in which the 
position of Vice President Plant Operations has been eliminated and the 
positions of Chief Operating Officer and Plant Manager were created. 
This change assigns certain management responsibilities to the Chief 
Operating Officer and Plant Manager.
    Date of issuance: August 1, 1996
    Effective date: August 1, 1996, to be implemented within 30 days of 
issuance.
    Amendment No.: 100
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25716) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 1, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 7th day of August 1966.
    For the Nuclear Regulatory Commission
Steven A. Varga, Director,
Division of Reactor Projects - I/II, Office of Nuclear Reactor 
Regulation
[Doc. 96-20586 Filed 8-13-96; 8:45 am]
BILLING CODE 7590-01-F