[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