[Federal Register Volume 61, Number 216 (Wednesday, November 6, 1996)]
[Notices]
[Pages 57481-57497]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-11106]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating
LicensesInvolving 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 October 11, 1996, through October 25, 1996.
The last biweekly notice was published on October 23, 1996.
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By December 6, 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
[[Page 57482]]
made a party to the proceeding; (2) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (3) the possible effect of any order which may be entered in the
proceeding on the petitioner's interest. The petition should also
identify the specific aspect(s) of the subject matter of the proceeding
as to which petitioner wishes to intervene. Any person who has filed a
petition for leave to intervene or who has been admitted as a party may
amend the petition without requesting leave of the Board up to 15 days
prior to the first prehearing conference scheduled in the proceeding,
but such an amended petition must satisfy the specificity requirements
described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: October 3, 1996
Description of amendment request: The proposed amendment changes
the provision for receiving, possessing and using byproducts, source
and special nuclear material at Calvert Cliffs Unit 1.
Currently, Unit 1 is licensed under 10 CFR Part 30 to receive,
possess, and use 100 millicuries of byproduct material for sample
analysis or instrument calibration, 500 millicuries of byproduct
material in the form of equipment; and 500 millicuries of Sodium-24 for
steam turbine acceptance testing. In addition, Unit 1 is licensed to
receive, possess and use 100 milligrams each of source or special
nuclear material under 10 CFR Parts 40 and 70. Unit 2 is licensed under
10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as
required any byproduct, source, or special nuclear material for sample
analysis or instrument calibration or associated with radioactive
apparatus or components. This proposed amendment would change the Unit
1 license to be consistent with the Unit 2 license by replacing license
conditions 2.B.3 and 2.B.4 with the same wording as Unit 2's license
condition 2.B.4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident.
Currently, Unit 1 is licensed under 10 CFR Part 30 to receive,
possess, and use 100 millicuries of byproduct material for sample
analysis or instrument calibration, 500 millicuries of byproduct
material in the form of equipment; and 500 millicuries of Sodium-24
for steam turbine acceptance testing. Unit 1 is also licensed under
10 CFR parts 40 and 70 to receive, possess, and use 100 milligrams
of source or special nuclear material. Unit 2 is licensed under 10
CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as
required any byproduct, source, or special nuclear material for
sample analysis or instrument calibration or associated with
radioactive apparatus or components. This proposed amendment would
change the Unit 1 license to be consistent with the Unit 2 license.
The reason for this proposed change is that it is sometimes
necessary to receive and use byproduct material, sources, or special
nuclear material with different activity levels, and in different
quantities than is specified by the Unit 1 license.
The current licenses for the two units allow radioactive
materials to be accepted and used at Unit 2, although these same
materials would not be acceptable for use at Unit 1. These
byproduct, source, and special nuclear materials are used by the
same people and for the same function in either
[[Page 57483]]
unit. Training and procedures for handling radioactive material have
been developed and used at both Units over the last 20 years. These
procedures are adequate to control the acceptance and use of
radioactive material at Unit 2 and, therefore, adequate to control
radioactive material at Unit 1.
Receiving, possessing, and using byproduct, source, or special
nuclear material is not related to accident conditions. Therefore,
changing the Unit 1 license conditions to be the same as the Unit 2
license condition does not involve a significant increase in the
probability or consequences of an accident.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
Procedures and training governing the acceptance and use of
radioactive materials are the same for both Unit 1 and Unit 2. These
procedures will not be changed as a result of this license change.
In addition, receiving, possessing, and using radioactive material
is not related to accident conditions. Therefore, making the Unit 1
license the same as the Unit 2 license will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in the margin of
safety.
The margin of safety in this case is exposure to contaminated
material or equipment. Exposure is controlled by adequate training
and procedures. Radioactive material is received by personnel
assigned to the Radiation Safety Section. These personnel are
trained in receiving and shipping contaminated material. Once the
material is onsite, it becomes the responsibility of the radiation
protection staff who are trained in the handling of all levels of
radioactive material. Training and procedures for handling
radioactive materials have been developed and used over the 20-year
life of the plant, and are currently deemed adequate for compliance
with the Unit 2 license. Therefore, making the Unit 1 license the
same as the Unit 2 license will not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: S. Singh Bajwa, Acting Director
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: September 20, 1996
Description of amendment request: The proposed amendments would add
a footnote to specification 4.3.1.B.4.A.10.a which refers to a letter
that describes enhancements made to the Combustion Engineering sleeve
installation process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment continues to allow the Combustion
Engineering sleeves to be used as an alternate tube repair method
for Zion steam generators, along with the process enhancements which
are described in the letter identified in the proposed Technical
Specification note. The sleeve configuration, which was designed and
analyzed in accordance with the criteria of Regulatory Guide (RG)
1.121 and Section III of the ASME Code, is unaffected by the
enhancements. Fatigue and stress analyses of the sleeved tube
assemblies as described in the currently approved Topical Report,
CEN-331-P, Revision 1-P, are unaffected by the enhancements.
Mechanical testing which has shown that the structural integrity
of the sleeves under normal, faulted, and upset conditions is within
the acceptable limits and is unaffected by the enhancements. Leakage
rate testing for the tube sleeves which has demonstrated that
primary to secondary leakage is not expected during any plant
condition is unaffected by the enhancements. The consequences of
leakage through the sleeved region of the tube, including the
enhancements, is bounded by the existing steam generator tube
rupture (SGTR) analysis included in the Zion Updated Final Safety
Analysis Report.
The proposed Technical Specification change reflects
enhancements to the installation and inspection process identified
in Topical Report CEN-331-P, Revision 1-P, which is currently
referenced in the Technical Specifications. These enhancements do
not increase the probability or consequences of an accident
previously evaluated. The enhancement which disallows the
installation of the tube plugs made from Inconel 600 material was
done so based upon industry information and is addressed by NRC
Bulletin 89-01. The use of the Plus Point Probe, its associated data
acquisition equipment, and improved visual inspection equipment, are
conservative actions and improve the quality of the sleeving
process. The use of the mechanical plug in lieu of the welded plug
meets the established design requirements and is advantageous in the
area of dose reduction, because of reduced time to install. Minor
changes to the sleeve installation equipment as described in the
Topical Report, represent equipment enhancements and do not alter
the sleeve design or qualification testing.
The proposed Technical Specification change does not adversely
impact any previously evaluated design basis accident. Installation
of the sleeves, with the described enhancements, can be used to
repair degraded tubes by returning the condition of the tubes to
their original design basis condition for tube integrity and leak
tightness during all plant conditions. Therefore, the currently
approved sleeving process with the described enhancements will not
increase the probability of occurrence of an accident previously
evaluated.
Therefore, these proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The implementation of the enhancements to the proposed sleeving
process will not affect the plant design basis. The current stress
and fatigue analyses of the repair identified in Topical Report CEN-
331-P, Revision 1-P, has shown the ASME Code and RG 1.121 allowable
values are met and are unaffected by the described enhancements. The
current sleeving design, with the described enhancements, will
continue to maintain overall tube bundle structural integrity and
leak tightness at a level consistent with that of the originally
supplied tubing. Leak and mechanical testing of the sleeves, are
unaffected by the proposed enhancements and continue to support the
conclusions that the sleeve retains both structural integrity and
leak tightness during all operating and accident conditions. Repair
of a tube with a sleeve, utilizing the described enhancements, does
not provide a mechanism that results in an accident outside of the
area affected by the sleeve.
The described change to implement the cited enhancements will
not create a new or different type of accident. The change only
reflects enhancements to the currently approved installation/
inspection process and, would not change or impact any hypothetical
accident previously discussed. Use of improved Non-Destructive
Examination, data acquisition and visual inspection equipment
improves the quality of the sleeving process and has no negative
effect on the margin of safety. The elimination of the use of the
Inconel 600 plug also improves the margin of safety.
Any hypothetical accident as a result of potential tube or
sleeve degradation in the repaired portion of the tube is bounded by
the existing SGTR analysis. The sleeve design, including described
enhancements, does not affect any other component, or affect any
location on the tube outside of the immediate area repaired.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
[[Page 57484]]
The currently approved sleeving repair of degraded steam
generator tubes has been shown by analysis to restore the integrity
of the tube bundle to its original design basis condition. By
implementing the described enhancements, the consistent quality of
the upper sleeve weld has increased thereby reducing the potential
for rework and reducing the potential for leaving a weld indication
in service.
The proposed change does not involve a reduction in the margin
of safety. The change reflects enhancements to the installation/
inspection processes which are currently referenced in the Technical
Specifications. These enhancements would not have any adverse
effects on the previously evaluated design transient or accident
analyses. The enhancements represent acceptable industry standards.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 21, 1996
Description of amendment request: The proposed amendment will
modify Containment Penetrations Nos. 53 and 65 design by modifying the
design of instrumentation lines for Containment Vacuum Relief (CVR)
system that pass through these containment penetrations. The proposed
change will correct the error in previously docketed information that
was used by NRC during licensing process.
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 change will not increase the probability of
previously analyzed accidents. The proposed change seeks to clearly
document the design and licensing bases for acceptance of the CVR
sensing instrument lines. The proposed change to the monitoring
lines will provide greater assurance that containment integrity will
be maintained following a LOCA concurrent with a single active
failure. The design change to the non-essential monitoring line will
reduce the potential bypass leakage from penetrations 53 and 65 by
adding a redundant automatic containment isolation valve on
penetration 53 and isolating the non-essential instrument line on
penetration 65. This design change can be performed at power without
violating any license/regulatory requirements that ensure
containment integrity is maintained.
There is no change in the function of the instrumentation. The
only difference is that CVR-IDPT-5017B and C non-safety differential
transmitters that monitor the CVR system will be sensing containment
pressure from penetration 53. If the non-essential line coming from
penetration 53 becomes inoperable, containment to annulus
differential pressure can be obtained from alternate
instrumentation. The essential sensing line that actuates the CVR
system to protect containment within design vacuum pressure is not
affected by the design change.
Adding a redundant automatic containment isolation valve in
penetration 53's non-essential instrument line instead of the excess
flow check valve and isolating the non-essential line in penetration
65's will significantly reduce the potential bypass leakage. The
proposed change will credit the essential instrument lines as a
closed system outside containment. The appropriate testing and
acceptance criteria will be applied to ensure that any leakage
associated with these potential bypass leakage paths, will not
exceed the limits used in the Waterford 3 safety analysis or result
in a significant increase in analyzed dose consequences. Therefore,
the proposed change will not involve significant increase in the
probability or consequences of any accident previously evaluated.
The proposed change will credit the essential sensing lines
outside containment as a closed system and will not affect the plant
or the manner in which the plant [is] operated.
The failure modes associated with containment isolation remain
unchanged as a result of the design change to the non-essential
monitoring lines. The function of the non-safety instrumentation is
not affected. The only difference is that all of the non-safety
instrumentation will be sensing containment pressure from
penetration 53. However, if the non-essential line coming from
penetration 53 becomes inoperable, containment pressure can be
obtained from alternate instrumentation. Adding a redundant
automatic containment isolation valve in series with CVR 401A in the
non-essential instrument line ensures containment isolation
following a LOCA with a concurrent a single active failure.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The addition of a redundant automatic containment isolation
valve in series with CVR 401A in the non-essential instrument line
breaching penetration 53 ensures containment isolation postulating a
single active failure on a Containment Isolation Actuation Signal
(CIAS). While this modification is performed, administrative
controls will require containment integrity to be maintained by a
seismic Category 1, ASME Section III, Class 2, passive containment
isolation device.
The essential CVR instrument sensing lines form a seismically
qualified, closed system outside containment which is designed for
pressure equal to or greater than containment. The instrument
cabinets C-3A(B) are seismic Category I and safety related. The
instruments are Safety Class 1E and have a static pressure rating of
1000 psig. These lines meet the criteria of BTP CSB 6-3 for
crediting a closed system as a leakage boundary to preclude bypass
leakage by being designed, fabricated, erected, and tested to
standards commensurate with the safety function to be performed. The
proposed change will apply the appropriate testing and acceptance
criteria to ensure that any leakage associated with these potential
bypass leakage paths, will not exceed the limits used in the
Waterford 3 safety analysis or result in a significant increase in
analyzed dose consequences. Therefore, the proposed change will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: October 4, 1996 (TSCR No. 250)
Description of amendment request: The proposed Technical
Specification (TS) change reflects a change in the Safety Limit Minimum
Critical Power Ratio (SLMCPR) and as a result, a change in the
operating Minimum Critical Power Ratio limit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The derivation of the revised SLMCPR for Oyster Creek for
incorporation into the TS,
[[Page 57485]]
and its use to determine cycle-specific thermal limits, have been
performed using NRC-approved methods. Additionally, interim
implementing procedures, which incorporate cycle-specific
parameters, have been used. Based on the use of these calculations,
the revised SLMCPR will not increase the probability or consequences
of an accident.
The basis of the MCPR Safety Limit calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid transition
boiling if the limit is not violated. The new SLMCPR preserves the
existing margin to transition boiling and fuel damage in the event
of a postulated accident. The probability of fuel damage is not
increased.
Revising the operating MCPR limit for stability will ensure that
adequate margin is retained to the SLMCPR.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit is a Technical Specification numerical
value designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident. The
stability MCPR limit ensures an adequate operating MCPR margin to
the SLMCPR. These revised limits cannot create the possibility of
any new type of accident. The new SLMCPR has been calculated using
NRC-approved methods. Additionally, interim procedures, which
incorporate cycle-specific parameters, have been used. Therefore,
the proposed TS change does not create the possibility of a new or
different kind of accident, from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new SLMCPR is calculated using NRC-approved methods which
are in accordance with the current fuel design and licensing
criteria. Additionally, interim implementing procedures, which
incorporate cycle-specific parameters, have been used. The MCPR
Safety Limit remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving fuel cladding integrity.
The revised stability MCPR limit retains the existing margin to the
SLMCPR. Therefore, the proposed TS change 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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: October 10, 1996 (TSCR No. 203)
Description of amendment request: The proposed Technical
Specification revision will extend the instrumentation surveillances
for Condenser Low Vacuum, High Temperature Main Steamline Tunnel,
Recirculation Flow, and Reactor Coolant Leakage. Additionally, the
change will extend the equipment tests/operability checks for
Containment Vent and Purge Isolation, Electromagnetic Relief Valve
Operability, and Drywell to Torus Leakage Test. The above change
extensions conform with the 24 month refueling interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated.
The proposed amendment extends the period between successive
refueling interval surveillance(s) to once every 24 months for those
surveillance(s) evaluated herein. The proposed surveillance interval
changes do not involve any change to the actual surveillance
requirements, nor does it involve any change to the limits and
restrictions on plant operations. The reliability of systems and
components relied upon to prevent or mitigate the consequences of
accidents previously evaluated is not degraded by the proposed
change to the surveillance interval. Assurance of system and
equipment availability is maintained. This change does not involve
any change to system or equipment configuration. Therefore, this
change does not increase the probability of occurrence or the
consequences of an accident previously evaluated.
Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment extends the period between successive
refueling interval surveillance(s) to once every 24 months for those
surveillance(s) evaluated herein. The proposed surveillance interval
changes do not involve any change to the actual surveillance
requirements, nor does it involve any change to the limits and
restrictions on plant operation. This change does not involve any
change to system or equipment configuration. Therefore, this change
is unrelated to the possibility of creating a new or different kind
of accident from any previously evaluated.
Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment extends the period between successive
refueling interval surveillance(s) to once every 24 months (+/-25%
or 30 months) for the surveillances evaluated herein. The proposed
surveillance interval changes do not involve any change to the
actual surveillance requirements, nor does it involve any change to
the limits and restrictions on plant operation. The reliability of
systems and components is not degraded by the proposed change to the
surveillance interval. Assurance of system and equipment
availability is maintained. Therefore, it is concluded that
operation of the facility in accordance with the proposed amendment
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: October 10, 1996 (TSCR No. 243)
Description of amendment: The proposed Technical Specification (TS)
will change the trip setting for TS Table 3.1.1 Item G.3, Automatic
Depressurization System (ADS) by clarification of the functional
requirement to provide an interlock permissive which ensures that a
source of cooling water is available via the Core Spray System prior to
depressurization. This will be accomplished by replacing the present
interlock description ``AC Voltage'' with core spray booster pump
differential pressure, as the permissive required for initiation of
ADS. A corresponding surveillance requirement is being added to TS
Table 4.1.1 which reflects the need to test and calibrate the core
spray booster pump differential pressure switches pursuant to existing
[[Page 57486]]
plant procedures. Additionally, allowed outage time (AOT) is addressed
in the footnote ``i'' for the differential pressure switches based upon
the currently designed ADS logic trains and footnote ``h'' to parallel
the ``Low-Low Reactor Water Level'' and ``High Drywell Pressure'' AOTs
associated with Standard Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated.
The implementation of this TSCR does not involve an increase in
the probability of occurrence or the consequences of an accident
previously evaluated, as no plant modifications are proposed by the
change request, and no changes in instrument set or reset setpoints
are required in order to implement the change. This change serves to
clarify and to incorporate the ``as-built'' ADS system logic
parameter (core spray booster pump differential pressure) as the
functional permissive required for initiation of ADS. This
``interlock'' permissive compares closely with that of the BWR
[boiling-water reactor] STS [Standard Technical Specifications]
requirement to monitor core spray discharge pressure for initiation
of ADS. In addition, the AOTs for the ADS initiation signals are
being revised to align with the AOTs provided for such signals in
the STS. The performance and function of the Automatic
Depressurization System is unchanged by this request. However, by
implementation of the change the specific functions of the ADS as-
built d/p permissives would then be clearly identified in and
controlled by T.S. Table 3.1.1, ``Protective Instrumentation
Requirements,'' including the associated surveillance requirements
as shown on the revised T.S. Table 4.1.1.
Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The implementation of this TSCR does not impact upon the
operation of the facility, and would not create the possibility of a
new or different kind of accident from any previously evaluated
because no plant modifications are proposed by this change request,
and no changes in instrument set or reset setpoints are required in
order to implement the change. This change clarifies the technical
specifications by incorporating the ``as-built'' ADS system logic
parameter (core spray booster pump differential pressure) as the
functional permissive required for initiation of ADS. This
``interlock'' permissive compares closely with that of the BWR STS
requirement to monitor core spay discharge pressure. The revised
AOTs for ADS initiation signals are also being changed to conform
with those allowed by and provided in the STS. The performance and
function of the Automatic Depressurization System (ADS) is unchanged
by this request.
OC plant surveillance procedures for both ADS and the Core Spray
system presently incorporate the calibration requirements and both
the set and reset setpoints calculated for the core spray booster
pump d/p switch permissive to the ADS initiation logic. Hence, a new
or different kind of accident from any previously evaluated is not
created.
Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The implementation of this TSCR does not involve a reduction in
the margin of safety for operation of the ADS or the Core Spray
system. The Technical Specification Bases which presently define the
margin of safety are not impacted as the core spray booster pump d/p
``interlock'' permissives are not described in the specifications
for ``Protective Instrumentation Requirements'' or its surveillance
requirements. In addition, the margin of safety for ADS initiation
is not reduced by this TSCR because the required system response is
not affected by the proposed changes as no plant modifications are
required which could create a potential impact upon the margins of
safety previously established.
The revision of AOTs associated with ADS actuation signals by
extension form 72 hours to 4 days is consistent with that presently
provided in the STS. This does not decrease the margin of safety
associated with availability of ADS as placement of the initiation
signals into the ``tripped condition'' maintains the operability of
the ADS trip systems while in the automatic mode. Additionally, the
Bases for STS Specification 3.1 provides justifications for AOTs
using the GE [General Electric] reliability analyses referenced
therein and therefore 4 days is both justified and conservative. The
margin of safety with respect to the instrument channels ability to
perform its intended actuation function is not impacted; therefore,
there is no reduction in the margin of safety.
Lastly, the surveillance frequency for the new surveillance
interval created on Table 4.1.1 for the d/p [s]witches is consistent
with that established in Reference 2 of the Bases for Technical
Specification 4.1. Therefore, there is no reduction in the margin of
safety as a result of this change request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: October 17, 1996
Description of amendment request: The proposed amendment would
revise Facility Operating License NPF-62 to acknowledge the transfer of
Soyland Power Cooperative's 13.21% minority ownership interest in the
Clinton Power Station to Illinova Power Marketing, Inc., the
unregulated power marketing affiliate of Illinois Power, and a wholly
owned subsidiary of Illinova Corporation.
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 because it merely revises the Operating License
to indicate the transfer of a minority ownership interest to the
corporate parent of the majority owner and licensee. This proposed
amendment represents an administrative rather than operational
change and, therefore, has no impact on accidents previously
evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated because Illinois Power will continue to be the operator of
Clinton
Power Station, and further, there will be no change to the
plant's physical configuration or operating philosophy as a result
of this proposed amendment.
3. The proposed amendment does not involve a significant
reduction in the margin of safety because it is only an
administrative change and will have no impact on any margin of
safety related to the design or operation of the facility.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, Illinois 62525
NRC Project Director: Gail H. Marcus
[[Page 57487]]
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: September 13, 1996
Description of amendment request: The proposed amendment would
revise the Maine Yankee containment testing technical specification (TS
4.4) to implement 10 CFR Part 50, Appendix J, Option B, by referring to
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test
Program'' dated September 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. This amendment request does not involve a significant
increase in the probability or consequences of an accident
previously evaluated, because the proposed changes to the Technical
Specification do not affect the assumption, parameters or results of
any FSAR accident analysis.[...] These changes potentially result in
a minor increase in the consequences of an accident previously
evaluated due to the increased testing intervals. However, the
proposed changes do not result in an increase in the probability of
an accident previously identified since the containment system is
used for mitigation purposes only. The changes are also expected to
result in increased attention to components with poor leakage test
history as part of the performance-based nature of Option B such
that the marginally increased consequences from the expanded testing
intervals may be further reduced or negated. The addition of the
''...[as modified by approved] exemptions'' phrase is an
administrative change. Any specific exemptions from the requirements
of Appendix J will continue to require a submittal under 10 CFR
50.12 and subsequent review and approval by the NRC prior to
implementation. Therefore, these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Operation of Maine Yankee in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) nor alter the
function of the containment system. The changes only provide for
additional time between leakage tests and an increase in the test
pressure value equal to the containment design pressure which bounds
the containment peak accident pressure. Thus, these changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Operation of Maine Yankee in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes do not alter the manner in which safety
limits, limiting safety system setpoints, or limiting conditions for
operation are determined. The changes are expected to result in an
increased focus on components demonstrating poor leakage test
history without excessive testing of components which continue to
demonstrate good test history. Therefore, these changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration. Local
Public Document Room location: Wiscasset Public Library, High Street,
P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:
John A. Zwolinski
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine
YankeeAtomic Power Station, Lincoln County, Maine
Date of amendment request: September 13, 1996, as supplemented
September 25, 1996
Description of amendment request: The proposed amendment would
revise TS 5.5.B to eliminate references to the Vice President (YNSD)
and designate the President, Maine Yankee, as the responsible official
for matters related to the composition, review and audit
responsibilities, authority and recordkeeping responsibilities of the
Nuclear Safety Audit and Review (NSAR) Committee. Minor editorial
changes are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below.
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change is administrative in nature and will not
have a direct effect on the physical plant or the maintenance of the
physical plant. The audit and review functions of the NSAR Committee
will continue to be required. The proposed changes will not, of
themselves, decrease the effectiveness of these functions. This
authority and responsibility realignment will continue to assure
that NSAR Committee has direct access to a level of management
necessary to perform their audit and review functions.
Since, the proposed change will not adversely effect the audit
and review functions of the NSARC and since the proposed change will
not have a direct effect on the physical plant or maintenance of the
physical plant, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change is administrative in nature and does not
introduce any new structures, systems, or components into the plant
design. This change continues to ensure that the NSAR Committee
reports to a management level such that there is sufficient
authority and organizational freedom to execute their audit and
review functions. Consequently, an unbiased oversight of the
programs and procedures is not compromised by this proposed change.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change realigns the authority and responsibility
relationship of the NSAR Committee. The NSAR Committee will continue
to maintain effective oversight of programs and procedures. The
proposed change will continue to ensure that the NSAR Committee is
sufficiently independent from cost and schedule when opposed to
safety considerations. Therefore, the proposed change does not
involve a significant reduction in the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:
John A. Zwolinski
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: July 2, 1996
Description of amendment request: The proposed amendment
incorporates limiting conditions for operation and surveillance
requirements for the safety/relief valve (SRV) electrical lift design
modification. The proposed amendment also makes clarification and
editorial changes, as well as revising the associated Bases section.
[[Page 57488]]
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:
Pursuant to 10CFR50.92, NNECO has reviewed the proposed change
and concludes that the change does not involve a significant hazards
consideration (SHC) since the proposed change satisfies the criteria
in 10 CFR 50.92(c). That is, the proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The safety relief valves are considered for two analyzed
accidents, an overpressure transient (such as MSIV [main steam
isolation valve] closure with flux scram) and an inadvertent SRV
opening.
The new technical specifications do not affect normal operation,
therefore, they cannot increase the probability of an overpressure
event. Since the mechanical function will not be affected by the new
equipment, the new LCOs [limiting conditions for operation], or the
new surveillance requirements, there is no adverse affect on the
consequences of an overpressure event. The SRVs will be expected to
lift mechanically. If they do not open at the design setpoints, the
electrical actuation, which has the same setpoints, will cause the
valves to open less than 400 milliseconds later.
Sufficient redundancy and diversity is established for the
electrical lift by the use of two sensors in a two-out-of-two-taken-
once configuration. Therefore, the failure of any single component
cannot result in an inadvertent opening of an SRV. The only proposed
surveillance performed while at power is the daily instrument check.
This surveillance does not require the manipulation of any controls
and, as such, cannot affect the probability of an accident.
Therefore, based on the above, the proposed change to the
Technical Specifications does not involve a significant increase in
the probability or consequences of any previously evaluated
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
None of the proposed new LCOs or surveillance requirements has a
potential for creating a new or different kind of accident.
Expanding the LCO and surveillance requirements to address both the
mechanical actuation and the pressure sensor lift does not change
the type of action that these valves are expected to perform, nor
does it change the initial ``as-left'' requirements for the valves.
Plant operating parameters have also not changed.
Therefore, this change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The margin of safety previously analyzed for the SRVs was based
on the current nominal setpoints and allowable percent drift. The
electrical lift system improves the confidence that the SRVs will
lift within the specified range. The setpoint uncertainty of the
electrical lift system is similar to the drift allowed for the
mechanical lift in the Technical Specifications. All existing
functions that may actuate the SRVs (safety, manual, or automatic
lift) remain unaffected. The design of the pressure transmitters,
combined with the logic configuration, minimizes the possibility of
inadvertently opening the SRVs.
Therefore, this change has no impact on 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270
NRC Project Director: Phillip F. McKee
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: March 29, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Surveillance Requirement
(SR) 4.5.1.d.2.b to delete the requirement to perform in-situ
functional testing of the Automatic Depressurization System (ADS)
valves once every 24-months as part of start-up testing activities.
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 Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS change does not involve any physical changes to
plant structures, systems, or components (SSC). The ADS will
continue to function as designed. The ADS is an Emergency Core
Cooling System (ECCS) designed to mitigate the consequences of an
accident, and therefore, can not contribute to the initiation of any
accident. The ADS utilizes five (5) of the 14 main steam line SRVs
as the primary method for depressurizing the reactor pressure vessel
to permit low pressure core cooling capability in the event of a
small break Loss-of-Coolant-Accident (LOCA) if the high pressure
cooling systems (i.e., High Pressure Coolant Injection (HPCI) and
Reactor Core Isolation Cooling (RCIC) systems) fail to maintain
adequate reactor vessel water level.
Deleting the TS SR to perform the in-situ testing of the ADS/
SRVs during start-up, as proposed, should reduce the probability of
an inadvertent opening of an SRV as discussed in Section 15.1.4 of
the LGS Updated Final Safety Analysis Report (UFSAR) since deleting
this testing requirement will eliminate a known initiator of SRV
pilot leakage and subsequent erosion. This proposed TS change will
have a tendency to increase, rather than decrease, the reliability
of the ADS/SRVs by eliminating the in-situ ADS functional start-up
testing. The probability of the ADS/SRVs to open on demand has been
demonstrated to be extremely high and is not measurably improved
through the in-situ ADS functional start-up testing.
This proposed TS change will not increase the probability of
occurrence of a malfunction of any plant equipment important to
safety. Alternate testing methods at LGS, Units 1 and 2, and at the
off-site test facility, adequately demonstrate proper ADS valve
operation and assure that the valves will continue to function as
designed. Existing surveillance testing and inspections of the ADS/
SRVs at LGS verify that the ADS initiation logic, solenoid valve
operation, pneumatic gas supply integrity and air operator assembly
(including pilot rod) will operate as designed. Offsite testing
verifies pilot disc operation, setpoint calibration and main valve
disc operation.
Deleting the in-situ testing requirement, as proposed, will
reduce the probability of inflating SRV leakage which should reduce
the probability of an inadvertent SRV opening. It has been
documented throughout the BWR industry that pilot disc leakage leads
to pilot disc and rod erosion, which can ultimately result in an
inadvertent opening of an SRV. Therefore, any SRV pilot leakage that
can be eliminated would reduce the probability of occurrence of a
malfunction of that SRV.
Deleting the ADS/SRV in-situ functional test will in no way
increase any consequences of a malfunction of plant equipment
important to safety. The consequences of a malfunction of an ADS/SRV
as discussed in the LGS UFSAR remain unchanged.
In addition, eliminating a known initiator of SRV leakage, as
proposed in this TS change, would help to reduce operator
workarounds in the form of suppression pool cooling and letdown
operation activities. As a result, this will reduce the unnecessary
operation of the Residual Heat Removal (RHR) and Residual Heat
Removal Service Water (RHRSW) systems.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
[[Page 57489]]
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed TS change does not involve any physical changes to
plant SSC. The design and operation of the ADS/SRVs is not changed
from that currently described in the Safety Analysis Report (SAR).
The ADS will continue to function as designed to mitigate the
consequences of an accident. No changes of any kind are being made
to the valves, auxiliary components, or ADS logic. Deleting the
requirement to perform the ADS in-situ functional test during plant
start-up as proposed in this TS Change Request reduces the
likelihood of a SRV developing a leak and degrading throughout the
subsequent operating cycle. There is no possibility that
implementing this proposed TS change would create a different type
of malfunction to the ADS/SRVs than any previously evaluated.
Eliminating the requirement to perform the in-situ testing of
the ADS/SRVs during start-up activities, does not create a new or
different type of accident than any previously evaluated. There is
no accident scenario associated with testing the ADS/SRVs other than
the inadvertent opening of a relief valve which is currently
discussed in Section 15.1.4 of the LGS UFSAR. This proposed TS
change does not alter the conclusions described in the UFSAR
regarding an inadvertent opening of an SRV. No new or different type
of accident will be created as a result of this proposed TS change.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed TS change does not involve any physical changes to
plant SSC. The design and functional requirements of the ADS will
not change. The ADS will still function as designed to mitigate the
consequences of an accident.
This proposed TS change involves deleting the requirement to
perform in-situ functional testing of the ADS/SRVs during start-up
activities. This testing imposes an unnecessary challenge on the
ADS/SRVs and has been linked to SRV degradation (e.g., pilot valve
and/or main valve leakage). This proposed TS change should reduce
SRV leakage and improve ADS/SRV reliability by reducing the
potential for spurious SRV actuation. The LGS TS Bases do not
identify specific testing requirements for ADS. ADS operability can
be readily demonstrated with extremely high confidence by the
existing additional surveillance tests and inspections performed for
the ADS. There will be no reduction in any margin of safety
resulting from this proposed TS change.
Therefore, the proposed TS 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: August 8, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Sections 3/4.3.1, ``Reactor
Protection System Instrumentation,'' 3/4.3.2, ``Isolation Actuation
Instrumentation,'' 3/4.3.3, ``Emergency Core Cooling System Actuation
Instrumentation,'' and the associated TS Bases Sections 3/4.3.1 and 3/
4.3.2 to eliminate selected response time testing requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes do not make any physical alterations or
modifications to the plant systems or equipment. The proposed
changes do not affect the capability of the associated systems to
perform their intended functions within their required response
times, nor do the proposed changes adversely impact the operation of
any plant equipment. The affected plant systems will continue to
function as designed. Elimination of the response time testing
requirements as proposed by this TS change for selected components
in RPS Instrumentation, Isolation Actuation System Instrumentation,
and ECCS Actuation Instrumentation will not adversely affect the
operation of these components.
The supporting analysis provided in NEDO-32291, demonstrates
that response time testing is redundant to other TS required
testing. NEDO-32291 demonstrated that these other required tests
(i.e., channel checks, channel calibrations, channel functional
tests, and logic system functional tests), in conjunction with
actions taken in response to NRC Bulletin 90-01 and NRCB 90-01,
Supplement 1, are sufficient to identify failure modes or
degradation in instrument response times, and ensure operation of
the associated systems within acceptable limits. There are no known
failure modes that can be detected by response time testing that
cannot also be detected by other TS required testing. The continued
application of other existing TS required testing such as channel
checks, channel calibrations, channel functional tests, and logic
system functional tests, ensures that the response times for these
systems will be maintained within the acceptance limits. The
capability of these systems to perform their intended functions
within their required response times is not adversely impacted by
this proposed TS change. NEDO-32291 evaluated the potential failure
modes of the affected instrumentation loops which could impact the
instrument loop response times. Industry operating experience was
also reviewed to identify failures that affect response times and
how they are detected. The failure modes identified were evaluated
to determine if other TS required surveillances and actions taken in
response to NRC Bulletin 90-01, and NRCB 90-01, Supplement 1, would
detect any effects on response time. There are no failures [sic]
[failure] modes identified that can be detected by response time
testing that cannot also be detected by other TS required testing.
PECO Energy has confirmed the applicability of the generic
evaluation provided in NEDO-32291 to LGS, Units 1 and 2. By letter
dated December 28, 1994, the NRC concluded that response time
testing can be eliminated from the TS for the selected
instrumentation identified in NEDO-32291, with certain provisions,
and that NEDO-32291 can be referenced in license amendment requests.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The proposed changes apply only to the
testing requirements for the selected components involved and do not
result in any physical modifications to these components, or to
other plant system components. Elimination of the response time
testing requirements as proposed by this TS change for selected
components in RPS Instrumentation, Isolation Actuation System
Instrumentation, and ECCS Actuation Instrumentation will not
adversely affect the operation of these components. These components
will continue to function as designed. Consequently, no new failure
modes are introduced as a result of the proposed TS changes.
Eliminating the response time testing requirements as proposed,
does not create a new or different type of accident than any
previously evaluated. No new or different type of accident will be
created as a result of this proposed TS change.
NEDO-32291 demonstrates that other required tests (i.e., channel
checks, channel calibrations, channel functional tests, and logic
system functional tests), in conjunction with actions taken in
response to NRC Bulletin 90-01 and NRCB 90-01, Supplement
[[Page 57490]]
1, are sufficient to identify failure modes or degradation in
instrument response times, and ensure operation of the associated
systems within acceptable limits. There are no known failure modes
that can be detected by response time testing that cannot also be
detected by other TS required testing, and therefore, response time
testing for the selected components is redundant to the other TS
required testing.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The proposed TS changes do not affect
the capability of the associated systems or equipment from
performing their intended functions. The systems involved will
continue to respond within their allowed response times. Elimination
of the response time testing requirements are based on the
evaluation provided in NEDO-32291 which demonstrates that response
time degradation can be detected by other TS required testing. The
evaluation concluded that other TS required tests (i.e., channel
checks, channel calibrations, channel functional tests, and logic
system functional tests), in conjunction with actions taken in
response to NRC Bulletin 90-01 and NRCB 90-01, Supplement 1, are
sufficient to identify failure modes or degradation in instrument
response times, and ensure operation of the associated systems
within acceptable limits.
In addition, although not specifically evaluated, the proposed
TS changes will provide an improvement to plant safety and operation
by reducing the time safety systems are unavailable, reducing the
potential for safety system actuations, reducing plant operating and
shutdown risk, limiting radiation exposure to plant personnel, and
eliminating the diversion of key personnel to conduct unnecessary
testing. Therefore, PECO Energy considers that the proposed TS
changes will result in an overall increase in the margin of safety
and that the changes do not constitute an unreviewed safety
question.
Therefore, the proposed TS changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket No. 50-353, Limerick
Generating Station, Unit 2, Montgomery County, Pennsylvania
Date of amendment request: August 1, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Section 3/4.4.6 (i.e.,
Figure 3.4.6.1-1) to reflect the addition of two hydrotest curves,
effective for 6.5 and 8.5 Effective Full Power Years (EFPY), to the
existing Pressure-Temperature Operating Limit (PTOL) curves for LGS
Unit 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed Technical Specification (TS) change includes
Pressure-Temperature Operating Limit (PTOL) curves which were
conservatively generated in accordance with the fracture toughness
requirements of 10CFR50, Appendix G. The Adjusted Reference
Temperatures to the initial nil ductility reference temperatures
(RTNDT) used to evaluate the pressure/temperature limits for the
beltline materials were based on Regulatory Guide 1.99, Revision 2.
Future analyses of the Reactor Pressure Vessel (RPV) surveillance
capsule contents and future revisions to the PTOL curve as required,
ensure that the reactor pressure boundary will behave in a non-
brittle manner during plant testing, startup, and operation
throughout the life of the plant. The current schedule for removal
of the surveillance specimens from Limerick Generating Station (LGS)
Unit 2 RPV is during 2R05. The proposed change does not impact the
existing PTOL curves for 10 Effective Full Power Years (EFPY),
currently shown in the LGS Unit 2 TS. The proposed change only
provides additional information (i.e., two new curves) related to
the RPV condition following 6.5 and 8.5 EFPY, in order to facilitate
hydrostatic testing performed after 2R04 and 2R05, respectively. The
added PTOL curves are established in compliance with the methodology
used to calculate the predicted irradiation effects on vessel
beltline materials as documented in the LGS Updated Final Safety
Analysis Report (UFSAR). There are no physical changes to the plant
being introduced by the added PTOL curves.
Therefore, the proposed (TS) change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed Technical Specification (TS) change includes
Pressure-Temperature Operating Limit (PTOL) curves which were
conservatively generated in accordance with the fracture toughness
requirements of 10CFR50, Appendix G. The Adjusted Reference
Temperatures to the initial nil ductility reference temperatures
(RTNDT) used to evaluate the pressure/temperature limits for the
beltline materials were based on Regulatory Guide 1.99, Revision 2.
The proposed changes do not impact the existing PTOL curves for 10
Effective Full Power Years (EFPY), currently shown in the TS. They
only provide additional information (i.e., two new curves) related
to the reactor pressure vessel condition for 6.5 and 8.5 EFPY, in
order to facilitate hydrostatic testing performed after 2R04 and
2R05, respectively. The added PTOL curves are established in
compliance with the previous methodology used to calculate the
predicted irradiation effects on vessel beltline materials as
documented in the LGS [Updated Final Safety Analysis Report] UFSAR.
The proposed TS change does not involve any physical changes to
safety-related equipment.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident, from any
accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed change to Technical Specifications (TS) does not
reduce the margin of safety as defined in the Bases for any TS. The
added Pressure-Temperature Operating Limit (PTOL) curves for 6.5 and
8.5 Effective Full Power Years (EFPY) corresponding to 2R04 and
2R05, respectively, have been calculated in accordance with the
existing methodology used to calculate the PTOL curves currently
existing in the LGS Unit 2 TS (i.e., complying with the requirements
of 10CFR50 Appendix G, and Regulatory Guide 1.99, Revision 2) and
will more closely reflect the actual required reactor pressure
vessel condition at the time in which the hydrotest is performed.
Therefore, the margin of safety is not affected.
Therefore, the proposed TS change 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
[[Page 57491]]
Philadelphia Electric Company, Docket No. 50-353, Limerick
Generating Station, Unit 2, Montgomery County, Pennsylvania
Date of amendment request: August 5, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Section 2.1 and its
associated TS Basis to reflect the change in the Minimum Critical Power
Ratio (MCPR) Safety Limit due to the plant specific evaluation
performed by General Electric Co. (GE), for LGS Unit 2 Cycle 4.
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 Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The revised Minimum Critical Power Ratio (MCPR) Safety Limit for
LGS Unit 2 Technical Specifications, and its use to determine cycle-
specific thermal limits have been performed using NRC-approved
methods within the existing design and licensing basis, and cannot
increase the probability or severity of an accident.
The basis of the MCPR Safety Limit calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid transition
boiling if the limit is not violated. The new MCPR Safety Limit
preserves the existing margin to transition boiling and fuel damage
in the event of a postulated accident. The probability of fuel
damage is not increased.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit is a Technical Specification numerical
value, designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident. It
cannot create the possibility of any new type of accident. The new
Minimum Critical Power Ratio (MCPR) Safety Limit is calculated using
NRC-approved methods and is based on LGS Unit 2 Cycle 4 specific
inputs.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident, from any
accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new Minimum Critical Power Ratio (MCPR) Safety Limit is
calculated using NRC approved methods which are in accordance with
the current fuel design and licensing criteria. The MCPR Safety
Limit remains high enough to ensure that greater than 99.9% of all
fuel rods in the core will avoid transition boiling if the limit is
not violated, thereby preserving the fuel cladding integrity.
Therefore, the proposed TS change 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: September 30, 1996
Description of amendments request: The proposed amendments would
revise Technical Specifications (TSs) 3/4.1.1, 3/4.1.3, 3.1.3.6, 3.2.1,
3/4.2.2, and 3.2.3 and associated Bases to remove certain cycle-
specific parameter limits from the TSs and relocate them to the Core
Operating Limits Report (COLR). These changes result from NRC Generic
Letter (GL) 88-16, dated October 4, 1988, which provided guidance to
licensees on requests for removal of the values of cycle-specific
parameter limits from the TSs. The licensee's proposed amendments are
consistent with the GL.
The COLR has been included in the Definitions section of the TSs.
The definition notes that it is the unit-specific document that
provides these limits for the current operating reload cycle. The
values of these cycle-specific parameter limits are to be determined in
accordance with TS 6.9.1.11. This TS requires that the core operating
limits be determined for each reload cycle in accordance with the
referenced NRC-approved methodology for these limits and consistent
with the applicable limits of the safety analysis. The COLR shall be
provided to the NRC upon issuance.In addition, the above TS changes
would produce administrative changes to the TS Table of Contents.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The removal of cycle-specific core operating limits from the FNP
[Farley Nuclear Plant] Technical Specifications has no influence or
impact on the probability or consequences of a Design Basis Accident
(DBA) occurrence. The cycle-specific core operating limits, although
not in Technical Specifications, will be followed in the operation
of FNP. The proposed amendment retains the same required actions to
be taken when or if limits are exceeded as stipulated by current
Technical Specifications. In addition, the associated surveillance
requirements are not altered by the proposed changes.
Each accident analysis addressed in the FNP FSAR [Final Safety
Analysis Report] will be examined with respect to changes in cycle-
dependent parameters, which are obtained from application of the
NRC-approved reload design methodologies, to ensure that the
transient evaluation of new reloads are bounded by previously
accepted analyses. This examination, which will be performed per
requirements of 10 CFR 50.59, ensures that future reloads will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As stated earlier, the removal of the cycle-specific variables
has no influence or impact, nor does it contribute in any way to the
probability or consequences of an accident. No safety-related
equipment, safety function, or plant operation will be altered as a
result of this proposed change. The cycle-specific variables are
calculated using the NRC-approved methods and submitted to the NRC
to allow the Staff to continue to trend the values of these limits.
The Technical Specifications will continue to require operation
within the required core operating limits and appropriate actions
will be taken when or if limits are exceeded. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed changes do not result in a significant reduction
in the margin of safety.
The margin of safety is not affected by the removal of cycle-
specific core operating limits from the Technical Specifications.
The margin of safety presently provided by current Technical
Specifications remains unchanged. Appropriate measures exist to
control the values of these cycle-specific limits. The proposed
amendment continues to require operation within the core limits, as
obtained from the NRC-approved reload design methodologies. The
required actions to be taken or if limits are violated remain
unchanged.
The development of the limits for future reloads will continue
to conform to those
[[Page 57492]]
methods described in NRC-approved documentation. In addition, each
future reload involves a 10 CFR 50.59 safety review to assure that
operation of FNP within the cycle-specific limits will not involve a
significant reduction in [the] margin of safety. Therefore, the
proposed changes are administrative in nature and do not impact the
operation of FNP in a manner that involves a reduction to 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: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
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.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
NuclearPower Station, Unit No. 1, New London County, Connecticut
Date of amendment request: August 29, 1996
Description of amendment request: The proposed amendment would
modify the applicability requirements for certain radiation monitors so
that the radiation monitors are required to be operable only when
secondary containment integrity is required to be operable; delineate
when secondary containment integrity is required; modify standby gas
treatment operability requirements; make editorial corrections to
clarify the configuration of the radiation monitors; and revise the
associated Bases section.
Date of publication of individual notice in Federal Register:
October 17, 1996 (61 FR 54242)
Expiration date of individual notice: November 18, 1996
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: September 6, 1996
Brief description of amendment request: The proposed amendment
would change Technical Specification (TS) requirements related to steam
generator tubes to allow a laser-welded repair of Westinghouse hybrid
expansion joint (HEJ) sleeved steam generator tubes. Date of individual
notice in Federal Register: October 15, 1996 (61 FR 53769)
Expiration date of individual notice: November 14, 1996
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 17, 1996
Brief description of amendments: The amendment modifies the
technical specifications (TS) to change (1) the reference method for
calculating dose conversion factors (DCFs) to be used in dose
calculations, and (2) the upper and lower limits for operating
pressurizer pressure to account for new instrument uncertainties and to
reduce the allowed operating band.
Date of issuance: October 23, 1996
Effective date: October 23, 1996, to be implemented within 45 days
of issuance
Amendment Nos.: Unit 1 - 109; Unit 2 - 101; Unit 3 - 81
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47963). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 23, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: June 9, 1995
[[Page 57493]]
Brief description of amendments: The amendments implement changes
to radiological effluent Technical Specifications in accordance with
Generic Letter 89-01 ``Implementation of Programmatic for Radiological
Effluent Technical Specification in the Administrative Controls Section
of the Technical Specifications and Relocation of Procedural Details of
RETS to the Offsite Dose Calculation Manual or to the Process Control
Program.''
Date of issuance: October 18, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 217 and 194
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 15, 1995 (60 FR
35062) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated October 18, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: May 1, 1996
Brief description of amendment: The proposed amendment will reflect
the implementation of 10 CFR Part 50 Appendix J, Option B at the
Pilgrim Nuclear Power Station.
Date of issuance: October 4, 1996
Effective date: October 4, 1996
Amendment No.: 167
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28606) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 4, 1996. No significant
hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
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, October 4, 1996, October 8, 1996, and October 16, 1996.
Brief description of amendment: The amendment changes the Technical
Specifications to (1) reflect the use of a new type of fuel (GE13) and
(2) modify the minimum critical power ratio safety limit and the
standby liquid control system sodium pentaborate limits to accommodate
the GE13 fuel.
Date of issuance: October 17, 1996
Effective date: October 17, 1996
Amendment No.: 182
Facility Operating License No. DPR-71: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42276) which superseded a Federal Register notice published on June 5,
1996 (61 FR 28607) The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated October 17, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of application for amendment: January 5, 1996, as supplemented
July 12, 1996
Brief description of amendment: The amendment revises the shutdown
cooling (SDC) requirement to allow one train of the SDC system to be
rendered inoperable for testing or maintenance provided that a filled
refueling cavity is available to provide backup decay heat removal
capability in the event that the operating train of SDC becomes
inoperable.
Date of issuance: October 10, 1996
Effective date: October 10, 1996
Amendment No.: 173
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44348) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 10, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: August 8, 1996
Brief description of amendments: The amendments revise the
Technical Specifications, Section 6.9.1.9, to reference updated or
recently approved topical reports used to calculate cycle-specific
limits contained in the Core Operating Limits Report.
Date of issuance: October 24, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 154 and 146
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47977) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 24, 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
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2, Pope County, Arkansas
Date of amendment request: May 9, 1996
Brief description of amendments: The amendments revised the name
from Arkansas Power & Light Company to Entergy Arkansas, Inc.
Date of issuance: October 23, 1996
Effective date: October 23, 1996
Amendment Nos.: 187 and 177
Facility Operating License Nos. DPR-51 and NPF-6. Amendments
revised the Technical Specifications and the licenses.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44357) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 23, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi,
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,
Claiborne County, Mississippi
Date of application for amendment: May 8, 1996, as supplemented by
letters dated July 18 and September 19, 1996
Brief description of amendment: The amendment modified the
frequency requirements in Surveillance Requirement 3.6.1.3.5 of the
Technical Specifications, on the leakage rate testing for each
containment purge
[[Page 57494]]
isolation valve with resilient seals, to place these purge valves on a
performance basis in accordance with Appendix J of 10 CFR Part 50, as
modified by any exemptions to Appendix J. In addition, the purge valves
would be required to be leakage rate tested every 36 months with at
least two pairs tested every 18 months and, if any purge valve fails to
meet the leakage rate acceptance criterion, all remaining valves must
be tested within 92 days (i.e., a quarter of a year) if not
successfully tested within the previous 92 days.
Date if issuance: October 18, 1996
Effective date: October 18, 1996
Amendment No.: 128
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28614) The additional information contained in the supplemental letters
dated July 18 and September 19, 1996, revised the proposed amendment in
the application of May 8, 1996; however, the revisions were within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 18, 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.
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: June 20, 1996, as supplemented
by the letter of September 11, 1996
Brief description of amendment: The amendment redefined the
secondary containment boundary to allow the enclosure building to be
inoperable during the upcoming refueling Outage 8 (RFO 8) scheduled to
begin in October 1996. The amendment added a condition to the license
that the enclosure building may be inoperable during core alterations
and movement of non-recently irradiated fuel (i.e., fuel that has not
occupied part of a critical reactor core for 12 days) during RFO 8 and
the standby gas treatment (SGT) system may be unable to automatically
start or achieve and maintain the required vacuum, provided the
following conditions exist:
a. All dampers communicating between the auxiliary building and the
enclosure building are closed.
b. The access door between the auxiliary building and the enclosure
building is closed, except when the access opening is being used for
entry and exit.
c. The SGT system is blocked from automatic initiation.
d. The SGT system is available for manual initiation or the actions
for Limiting Condition for Operation 3.6.4.3 in the Technical
Specifications for GGNS are complied with.
The non-recently irradiated fuel is spent fuel that has decayed at
least 12 days after the reactor was shut down for refueling.
Date of issuance: October 18, 1996
Effective date: October 18, 1996
Amendment No: 129
Facility Operating License No. NPF-29. Amendment adds a condition
to the license.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37299) The additional information contained in the supplemental letter
of September 11, 1996, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated October 18, 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.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: February 22, 1996, and as
supplemented by letters dated July 22 and September 20, 1996
Brief description of amendment: The amendment revises Clinton Power
Station Technical Specification 3.4.11, ``Reactor Coolant System (RCS)
Pressure and Temperature (P/T) Limits,'' to incorporate specific P/T
limits for the bottom head region of the reactor vessel, separate and
apart from the core beltline region of the reactor vessel.
Date of issuance: October 23, 1996
Effective date: October 23, 1996
Amendment No.: 109
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18169) The letters of July 22 and September 20, 1996, provided
clarifying information and did not alter the staff's initial finding
that the proposed changes involve no significant hazards consideration.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 23, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: November 21, 1995
Brief description of amendment: The amendment changes Technical
Specification Section 5.2.2, ``Design Pressure and Temperature,'' to
clarify that the reactor containment design temperature is an
equilibrium liner temperature and not the air temperature. The
supporting Technical Specification Bases is updated to reflect the
change and to include the main steam line break accident, in addition
to the loss-of-coolant accident, as the limiting events affecting the
containment temperature and pressure.
Date of issuance: October 21, 1996
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 204
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65684) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 21, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment requests: July 15, 1996, as supplemented by
letters dated September 3, 1996, October 22, 1996, October 23, 1996,
and August 23, 1996
Brief description of amendment: The amendment revises Technical
Specifications (TS) Section 4.3.2 to allow the use of zircaloy or ZIRLO
fuel
[[Page 57495]]
cladding and to use depleted uranium as reactor fuel material. The
amendment also changes TS Section 5.9.5 to add Westinghouse Topical
Reports, WCAP-12610-P-A, ``VANTAGE + Fuel Assembly Report,'' and WCAP-
13027-P, ``Westinghouse ECCS Evaluation Model for Analysis of CE-
NSSS,'' to the list of approved analytical methods for determining the
core operating limits.
Date of issuance: October 25, 1996
Effective date: October 25, 1996
Amendment No.: 178
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40026) and August 30, 1996 (61 FR 45995). The September 3, 1996,
October 22, 1996, and October 24, 1996, supplemental letters provided
additional clarifying and correcting information and did not change the
initial no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 25, 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: June 7, 1996
Brief description of amendments: The amendments revised the
combined Technical Specifications (TS) for the Diablo Canyon Nuclear
Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications 3/
4.9.14.1, ``Spent Fuel Assembly Storage - Spent Fuel Pool Region 2,''
and TS 3/4.9.14.3, ``Spent Fuel Assembly Storage - Spent Fuel Pool
Region 1,'' to allow storage of fuel assemblies in a checkerboard
pattern in Region 2 of the spent fuel pool (SFP).
Date of issuance: October 25, 1996
Effective date: October 25, 1996, to be implemented within 30 days
from date of issuance.
Amendment Nos.: Unit 1 - 116; Unit 2 - 114
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 25, 1996 (61
FR 50346) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 25, 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
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: December 19, 1995, as
supplemented by letter dated August 8, 1996.
Brief description of amendments: The amendments revised the
combined Technical Specifications (TS) for the Diablo Canyon Power
Plant Unit Nos. 1 and 2 to relocate Technical Specification (TS) 6.5,
``Review and Audit,'' 6.8, ``Procedures and Programs,'' Sections
6.8.1c., 6.8.1d., 6.8.2, and 6.8.3, in accordance with guidance in an
NRC letter dated October 25, 1993, from William T. Russell to the
chairpersons of industry owners groups and the Commission's Final
Policy Statement on TS Improvements for Nuclear Power Reactors on
relocation of TS that do not satisfy the retention criteria. As part of
the relocation of TS 6.8.2, TS 6.1.1 would be revised to require that
proposed tests, experiments, or modifications that affect nuclear
safety be approved by the plant manager or his designee prior to
implementation.
Date of issuance: October 25, 1996
Effective date: October 25, 1996, to be implemented within 90 days
of issuance.
Amendment Nos.: Unit 1 - 117; Unit 2 - 115
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1633) The August 8, 1996, supplemental letter provided additional
clarifying information and did not change the staff's initial no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 25, 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
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 application for amendments: May 20 and 28, 1996, as
supplemented by letter dated July 25, 1996
Brief description of amendments: These amendments, for both units,
add a reference to the ANF-B critical power correlation to Section
6.9.3.2 of the Technical Specifications (TSs); change the values of the
minimum critical power ratio (MCPR) in TS Sections 2.1 and 3.4.1.1.2,
and make appropriate Bases changes. For Unit 1 only, a reference to ABB
licensing methodology report CENPD-300 (for lead use assemblies being
used in the reactor core during the upcoming operating cycle) is added
to Section 6.9.3.2.
Date of issuance: October 11, 1996
Effective date: For both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 161 and 132
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: Unit 2, August 28, 1996
(61 FR 44362); Unit 1, September 4, 1996 (61 FR 47529)The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 11, 1996.No significant hazards consideration
comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: March 29, 1996, as supplemented
July 12, 1996, and September 6, 1996.
Brief description of amendment: The proposed amendment would change
the Indian Point 3 Technical Specifications (TSs) 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
rods. The proposed amendment would also add limits associated with
Departure from Nucleate Boiling to the IP3 Technical Specifications
TSs.
Date of issuance: October 22, 1996
Effective date: As of the date of issuance to be implemented within
30 days
[[Page 57496]]
Amendment No.: 170
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 17, 1996 (61 FR
37301) August 14, 1996 (61 FR 42283)The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated October 22,
1996.No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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: July 19, 1995, as supplemented
by letters dated December 22, 1995, and March 26, 1996.
Brief description of amendments: These amendments modify Technical
Specification (TS) 3.3.8, ``Containment Purge Isolation Signal
(CPIS),'' and TS 3.3.9, ``Control Room Isolation Signal (CRIS).'' The
revisions are needed to (1) support the upgrading or replacement of
existing radiation monitoring system with state-of-the-art equipment
that will provide for greater operational flexibility and reliability,
and (2) incorporate minor editorial changes to improve clarity of these
TS sections.
Date of issuance: October 8, 1996
Effective date: October 8, 1996, to be implemented within 30 days
of date of issuance
Amendment Nos.: Unit 2 - Amendment No. 132; Unit 3 - Amendment No.
121
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49948). The December 22, 1995, and March 26, 1996, letters provided
additional clarifying information and did not change the initial no
significant hazards consideration determination.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated October 8, 1996.No significant hazards consideration
comments received: No.
Temporary Local Public Document Room location: Science Library,
University of California, P. O. Box 19557, Irvine, California 92713
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: March 29, 1996, as supplemented by
letters dated June 27, August 29, and September 16, 1996.
Brief description of amendment: The amendment changes Technical
Specification 3/4.4.6, ``Steam Generators'' and associated Bases to
modify the steam generator repair limit to clarify that the appropriate
method for determining serviceability for tubes with outside diameter
stress corrosion cracking at the tube support plate is by a methodology
that more reliably assesses structural integrity.
Date of issuance: October 11, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment No.: 115
Facility Operating License No. NPF-8: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25711) The June 27, August 29, and September 16, 1996, letters provided
additional, clarifying information that did not change the scope of the
March 29, 1996, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 11, 1996. No significant hazards
consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: April 22, 1996, as supplemented by
letters dated May 3, July 15, August 7 and 30, and September 16, 1996
Brief description of amendment: The amendment changes reflect the
implementation of a new F* criterion based on maintaining existing
safety margins for steam generator tube structural integrity concurrent
with allowances for nondestructive examination eddy current
uncertainty.
Date of issuance: October 11, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment No.: 116
Facility Operating License No. NPF-8: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25713) The May 3, July 15, August 7 and 30, and September 16, 1996,
letters provided clarifying information that did not change the scope
of the April 22, 1996, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 11, 1996. No significant hazards
consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of application for amendment: June 29, 1996
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 5.2.2.f to delete the sentence, ``The
Operations Manager shall hold or have held an SRO [Senior Reactor
Operator] license on a similar unit.'' The revision also indicates that
the Operations Superintendent will hold a valid SRO license on this
unit.
Date of issuance: October 15, 1996
Effective date: Octber 15, 1996
Amendment No.: 4
Facility Operating License No. NPF-90: Amendment revises the TS.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47983)The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 15, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: July 31, 1996 (TXX-96432) as
supplemented by letters dated August 23 and 27 (TXX-96447 and TXX-
96451), and September 19, 1996 (TXX-96469).
Brief description of amendments: The amendments (1) change the
acceptance values for amperes and voltage for the 18 month surveillance
test of the battery chargers; (2) clarify the meaning of the
[[Page 57497]]
term ``associated inverter'' used in the context of energizing 118-Volt
AC Instrument Buses during MODES 1 through 6; and (3) delete the
protection channel and the vital bus ratings for the 118-Volt AC
Instrument Buses identified for MODES 1 through 4.
Date of issuance: October 22, 1996
Effective date: October 22, 1996
Amendment Nos.: Unit 1 - Amendment No. 53; Unit 2 - Amendment No.
39
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44363) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 22, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Dated at Rockville, Maryland, this 30th day of October 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II,Office of Nuclear Reactor
Regulation
[Doc. 96-28372 Filed 11-5-96; 8:45 am]
BILLING CODE 7590-01-F
