[Federal Register Volume 60, Number 216 (Wednesday, November 8, 1995)]
[Notices]
[Pages 56361-56378]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-11108]
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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
[[Page 56362]]
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 14, 1995, through October 27, 1995.
The last biweekly notice was published on Wednesday, October 25, 1995
(60 FR 54714).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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 8, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no
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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, 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, 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-529 and
STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 2 and
3, Maricopa County, Arizona
Date of amendments request: October 3, 1995
Description of amendments request: The amendment would delete the
provisions relating to certain previous sale and leaseback transactions
that were by added by Amendment No. 3 for NPF-51 and Amendment No. 1
for NPF-74.
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.
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed change is administrative in nature.
The proposed change deletes Sections 2.B.(7)(a) and (b) of License
No. NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74.
These sections describe the structure of the financing of El Paso's
interest in Palo Verde, specifically authorizing sale and leaseback
transactions. The proposed change does not affect the assumptions
used in the accident
analyses, nor does the proposed change result in changes to the
physical configuration of the facility, design parameters, technical
specifications, or operation and maintenance of the facility.
Therefore, the amendment request 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.
This amendment request does not create the possibility of a new
or different kind of accident from any accident previously analyzed
because the proposed change is administrative in nature. The
proposed change deletes Sections 2.B.(7)(a) and (b) of License No.
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These
sections describe the structure of the financing of El Paso's
interest in Palo Verde Units 2 and 3, specifically authorizing sale
and leaseback transitions. The proposed change does not involve
modifications to any of the existing equipment nor does the change
affect operation or maintenance of the facility. Therefore, the
amendment request does not create the possibility of a new or
different kind of accident not previously analyzed.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This amendment request does not involve a significant reduction
in a margin of safety because it is administrative in nature. The
proposed change deletes Sections 2.B.(7)(a) and (b) of License No.
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These
sections describe the structure of the financing of El Paso's
interest in Palo Verde, specifically authorizing the sale and
leaseback transactions. The proposed change does not involve changes
to any existing plant equipment or accident analyses that provide
for or establish margins of safety. There is no change to the
operation or maintenance of the facility and the existing margins of
safety are not changed by the proposed change. Therefore, the
amendment request does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Baltimore Gas and Electric Company, Docket No. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of amendment request: October 2, 1995
Description of amendment request: The proposed amendment would
revise the Calvert Cliffs Nuclear Power Plant, Unit No. 2, Technical
Specifications on a one-time basis by increasing the 7 day allowed
outage time (AOT) of the control room emergency ventilation system
(CREVS) to an AOT of 30 days. This requested one-time increase in the
AOT is applicable only for the loss of the emergency power supply to
one train of the CREVS during the Unit No. 1 spring 1996 refueling
outage.
The requested extension in the AOT is necessary to allow the
licensee to perform modifications to the electrical distribution system
during the upcoming Unit 1 refueling outage while Unit No. 2 continues
to operate. The modifications include connecting a fourth safety-
related (SR) emergency diesel generator (EDG) to engineered safety
features (ESF) Bus No. 11. The work related to this effort will require
that the bus be deenergized for several days isolating it from its
normal and
[[Page 56364]]
emergency EDG power supplies. One train of the CREVS is connected to
ESF Bus No. 11 and will not have its power supplies available for a
period of time. The normal (offsite) power is expected to be restored
in about 3 days, but the emergency power (onsite EDG) may take up to 30
days.
The licensee is taking additional actions to assure the
availability of the normal offsite power source and is also adding a
nonsafety-related (NSR) EDG as an alternate onsite power source during
the period that the SR EDG is not available. The licensee expects that
the tie-in of the NSR EDG will take about 8 days. Thus, even if the
normal offsite power source is lost, the temporary onsite NSR EDG will
be available to provide power to the affected train of the CREVS.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Control Room Emergency Ventilation System (CREVS) is used to
mitigate the consequences of an accident. It is designed so that the
Control Room remains habitable for operators and to maintain the
environment needed for continued equipment operation. The system is
redundant (two 100% capacity trains) and is powered from both normal
(offsite) and emergency (emergency diesel generators) power sources.
We [the licensee] are proposing an amendment which would allow the
emergency power to be removed from one of the redundant CREVS for an
additional 23 days (beyond the 7 days allowed by the Technical
Specifications). Other than the removal of the emergency electrical
power source, we are not affecting or modifying the operation of the
CREVS. The CREVS is not an accident initiator for any previously
evaluated accident. Therefore, the proposed change does not involve
an increase in the probability of an accident previously evaluated.
The CREVS is designed to mitigate the consequences of design
basis accidents. For that purpose, redundant trains are provided to
protect against a single failure. During the Technical Specification
seven day Allowed Outage Time (AOT), an operating unit is allowed by
the Technical Specifications to remove one of the CREVS trains from
service, thereby eliminating this single failure protection. The
consequences of a design basis accident coincident with a failure of
the redundant CREVS train during the additional 23-day period are
the same as those during the 7-day AOT. Therefore, the proposed
change does not significantly increase the consequences of an
accident previously evaluated.
Therefore, the proposed change does not increase the probability
or consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The CREVS is not being modified by this proposed change nor will
any unusual operator actions be required. The system will continue
to operate in the same manner. The CREVS is not an initiator to any
accident, but is designed to respond should an accident occur.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The operability of the CREVS during Modes 1 through 4 ensures
that the Control Room will remain habitable for operators and to
maintain the environment needed for continued equipment operation
under all plant conditions. The proposed change does not affect the
function of the CREVS. During the period of the Technical
Specifications AOT when one CREVS train is inoperable, the margin of
safety is reduced. This time period is a temporary relaxation of the
single failure criteria, which, consistent with overall system
reliability considerations, provides a limited time to maintain or
repair the equipment and conduct testing. We are requesting an
extension of this limited time. The proposed change will allow one
train of the CREVS to be without an emergency power supply for an
additional 23 days beyond the 7-day AOT (total of 30 days). This
train of CREVS will be functional and will have the normal power
supply available for all but approximately three days to allow work
and necessary testing on the bus. The other train of the CREVS will
have both its normal and emergency power supplies during this
period.
To provide additional assurance that all reasonable steps have
been taken to prevent the loss of the normal power supply to the
CREVS, we will restrict maintenance activities on three of the four
offsite transmission lines. This restriction will cover the period
we are in the Action Statement for the CREVS (Action Statement
3.7.6.1.a and b). To provide an alternative power source during the
majority of this period, we will connect the Alternate AC power
source (No. 0C Diesel Generator) to ESF Bus No. 11 and confirm its
availability as soon as possible after the work on ESF Bus No. 11
begins (we [the licensee] expect that to take about eight days).
This power source is independent from the offsite power supplies. In
addition, we will restrict planned maintenance on the No. 12 CREVS
during the period we are in the Action Statement to ensure that the
No. 12 CREVS is not removed from service.
We believe that the reduction in the margin of safety
represented by this one-time extension of the AOT is not significant
based on our management of plant risk, the reliability of the normal
CREVS power supply, the availability of the redundant CREVS with
both its normal and emergency power, and the mitigating features
described above. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment 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: Ledyard B. Marsh
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County, North Carolina
Date of amendments request: October 23, 1995
Description of amendments request: The amendments would delete the
applicability of the primary coolant water chemistry limits when the
primary system is being chemically decontaminated and the reactor
vessel is defueled.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed changes will allow the reactor coolant system
conductivity and chlorides to exceed the limits specified in
Technical Specification Table 3.4.4-1 in support of performing
chemical decontamination activities. The reactor coolant system
water chemistry limits have been established to prevent long-term
damage to the reactor coolant system materials that are in contact
with the coolant. Upon concluding the chemical decontamination
activities, reactor coolant system conductivity and chloride values
would be restored to within the limits specified in Technical
Specification Table 3.4.4-1. Existing regulatory requirements,
specifically a review in accordance with 10 CFR 50.59 to determine
whether an activity involves an unreviewed safety question, provide
adequate assurance that solvents selected for use in a chemical
decontamination activity will not degrade the structural integrity
of the reactor coolant system. Therefore, since the structural
integrity of the reactor coolant system will not be adversely
impacted by the chemical decontamination activities, the proposed
amendments do not involve a significant increase in the probability
of an accident previously evaluated.
As discussed above, the reactor coolant system water chemistry
limits have been
[[Page 56365]]
established to prevent long-term damage to the reactor coolant system
materials that are in contact with the coolant. The solvents being
used for a chemical decontamination activity are selected to ensure
their effectiveness and to ensure that damage will not occur to the
structural materials comprising the reactor coolant pressure
boundary. As such, the operation of safety equipment used to
mitigate a design basis accident or transient will not be affected
by the proposed change of the reactor coolant system water chemistry
limits during performance of chemical decontamination activities.
Therefore, the proposed revision to the reactor coolant system
chemistry limits will not involve a significant increase in the
consequences of an accident previously evaluated.
2. The proposed change will allow the reactor coolant system
conductivity and chlorides to exceed the limits specified in
Technical Specification Table 3.4.4-1 in order to perform chemical
decontamination activities. The reactor coolant system water
chemistry limits have been established to prevent long-term damage
to the reactor coolant system materials that are in contact with the
coolant. Even though the solvents used for chemical decontaminations
may result in reactor coolant system conductivity and chloride
measurement values in excess of the limits specified in the
Technical Specifications, the existing regulatory requirements of 10
CFR 50.59 will continue to ensure that solvents being used for
performing chemical decontamination have been properly evaluated and
that these solvents do not adversely affect the material properties
or structural integrity of the reactor coolant system. Therefore,
the proposed amendments revising the reactor coolant system water
chemistry limits during performance of chemical decontamination
activities will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The reactor coolant system water chemistry limits have been
established to prevent long-term damage to the reactor coolant
system materials that are in contact with the coolant. The solvents
used for chemical decontaminations result in reactor coolant system
conductivity and chloride measurement values in excess of the limits
specified in the Technical Specifications; however, the solvents
being used of performing chemical decontamination have been properly
evaluated to ensure they will not significantly affect the material
properties of the reactor coolant system piping (i.e., corrosion)
nor will they significantly affect the structural integrity (i.e.,
wall thinning) of the reactor coolant system piping. Therefore, the
proposed license amendments 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: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: November 2, 1994, as supplemented on
January 4, 1995
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to make editorial changes, delete
portions of the TSs that have become unnecessary due to previously
approved amendments, change managerial titles, update references and
reporting requirements, revise the Station Nuclear Safety Committee
(SNSC) composition to specify disciplines rather than specific job
titles, modify the record keeping requirements of the Nuclear
Facilities Safety Committee, implement changes referenced in Generic
Letter 93-07, ``Modification of the Technical Specification
Administrative Control Requirements for Emergency and Security Plans,''
and to correct the shift manning requirements table.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. There is no significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendments are administrative in nature. They
involve making editorial changes, deleting portions of the Technical
Specifications that have become unnecessary due to previously
approved amendments, changing managerial titles, updating references
and reporting requirements, revising the SNSC composition to specify
disciplines rather than specific job titles, implementing changes
referenced in Generic Letter 93-07, and revising shift manning to
conform with the requirements of 10 CFR 50.54. These changes do not
affect possible initiating events for accidents previously evaluated
or alter the configuration or operation of the facility. The
Limiting Safety Systems Settings and Safety Limits specified in the
current Technical Specifications remain unchanged. Therefore, the
proposed changes to the subject Technical Specifications would not
increase the probability or consequences of an accident previously
evaluated.
2. The possibility of a new or different kind of accident from
any accident previously evaluated has not been created.
As stated above, the proposed changes are administrative in
nature. The safety analysis of the facility remains complete and
accurate. There are no physical changes to the facility and the
plant conditions for which the design basis accidents have been
evaluated are still valid. The operating procedures and emergency
procedures are unaffected. Consequently, no new failure modes are
introduced as a result of the proposed changes. Therefore, the
proposed changes would not initiate any new or different kind of
accident.
3. There has been no significant reduction in the margin of
safety.
The proposed changes are administrative in nature. Since there
are no changes to the physical design or operation of the facility,
the Updated Final Safety Analysis Report (UFSAR) design basis,
accident assumptions, or Technical Specification Bases are not
affected. Therefore, the proposed changes would not result in a
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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003
NRC Project Director: Ledyard B. Marsh
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: August 29, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification Sections 3.1.F and 4.13 to provide for
appropriate inservice inspection for any steam generator tubes
containing sleeves and to provide for reduced allowable primary-to-
secondary leakage rates for steam generators containing sleeves. The
proposed changes are in response to commitments made by Consolidated
Edison by letter dated April 5, 1995, during the review of an amendment
which permitted the use of laser welded steam generator tube sleeves as
a method of tube repair.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the
[[Page 56366]]
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.
Technical Specification Amendment No. 183 allowed sleeving as an
acceptable alternate tube repair method for Indian Point Unit No. 2.
The steam generator sleeve approved for installation is the
Westinghouse process (laser welded sleeve). The sleeve configuration
was designed and analyzed in accordance with the criteria of
Regulatory Guide (RG) 1.121 and the design requirements of Section
III of the American Society of Mechanical Engineers (ASME) Code.
Fatigue and stress analyses of the sleeved tube assembly produced
acceptable results as documented in the Westinghouse topical report
submitted in the original sleeving package. Mechanical testing has
shown that the structural strength of the sleeves under normal,
faulted, and upset conditions is within acceptable limits. Leakage
rate testing for the tube sleeves has demonstrated that primary-to-
secondary leakage is not expected during all plant conditions.
Any leakage through the sleeved region of the tube is fully
bounded by the leak-before-break considerations and, ultimately, the
existing steam generator tube rupture analysis included in the
Updated Final Safety Analysis Report (UFSAR).
The reduction in TS leakage rate requirements from 0.3 gpm
[gallons per minute] (432 gpd [gallons per day]) allowable per SG to
150 gpd per steam generator containing sleeves further ensures that
SG tube integrity is maintained in the event of a main steam line
break (MSLB) or under Loss Of Coolant Accident (LOCA) conditions.
The RG 1.121 criteria for establishing operational leakage rate
limits require a plant shutdown based upon a leak-before-break
consideration to detect a free span crack before a potential tube
rupture. The 150 gpd limit will continue to allow for early leakage
detection and require a plant shutdown in the event of tube leakage
that exceeds the revised Technical Specification limit.
The sleeve sample size has been increased to a minimum of twenty
(20) percent of the inservice sleeves. Increasing the sample size of
the sleeves to be inspected will increase the monitoring of tubes
using sleeves for any further degradation while they remain
inservice. If the sample identifies a sleeve with an imperfection of
greater than 23 percent depth an additional 20 percent of the
sleeves shall be inspected. The sleeves that have identified
imperfections of greater than 23 percent shall be evaluated and
removed from service.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Implementation of the proposed amendment will not introduce
significant or adverse changes to the plant design basis. The
proposed changes do not involve plant modification or changes to
equipment, and consist of reducing the allowable steam generator
leakage limits for steam generators containing sleeves and defining
the sample size of the steam generator tube sleeve inspection.
The reduction in TS leakage rate requirements from 0.3 gpm (432
gpd) allowable per SG to 150 gpd per SG containing sleeves further
ensures that SG tube integrity is maintained in the event of a MSLB
or under LOCA conditions. The 150 gpd limit is designed to provide
for leakage detection and a plant shutdown in the event of the
concurrence of excessive tube leakage. The limit provides for early
detection and a plant shutdown prior to a postulated defect reaching
critical magnitudes for Main Steam Line Break conditions.
Formalizing the sample size of sleeved tubes inspected during
each scheduled inservice inspection will ensure increased monitoring
of these tubes for any further degradation. The improved monitoring
and evaluation of the tube and the sleeves assures tube structural
integrity is maintained or the tube is removed from service.
With these actions the possibility of a new or different type of
accident from any accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Implementation of the proposed changes will not reduce the
margin of safety. This amendment involves the reduction of sleeved
steam generator tube leakage limit and a formalized inservice
inspection program for sleeved tubes. These actions will help ensure
steam generator tube integrity.
Reduction of the leakage rate requirement from 0.3 gpm (432 gpd)
to 150 gallons per day (gpd) per sleeved steam generator will
continue to ensure steam generator tube integrity is maintained in
the event of main steam line break or under LOCA conditions.
Reducing this limit will not result in a reduction in the margin of
safety.
The portions of the installed sleeve assembly which represent
the reactor coolant pressure boundary will be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the requirement of Regulatory Guide 1.83. The portion of
the tube bridged by the sleeve joints is effectively removed from
the pressure boundary, and the sleeve then forms the new pressure
boundary. The sleeve enhances the safety of the plant by increasing
the protective boundaries of the steam generator. Keeping the tube
in service with the use of a sleeve, instead of plugging the tube
and removing it from service, increases the heat transfer efficiency
of the steam generator. Monitoring for any increased degradation of
a repaired steam generator tube shall be implemented by sampling
twenty (20) percent of the sleeves inservice. During each scheduled
inservice inspection, any sampled sleeve evaluated and found to have
unacceptable degradation shall be removed from service.
Based on the preceding analysis it is concluded that operation
of Indian Point Unit No. 2 in accordance with the proposed amendment
does not increase the probability of an accident previously
evaluated, does not create the possibility of a new or different
kind of accident from any accident previously evaluated, nor reduce
any margin of plant safety. Therefore, the license amendment does
not involve a Significant Hazards Consideration as defined in 10 CFR
50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003
NRC Project Director: Ledyard B. Marsh
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: September 30, 1994, as supplemented by
letter dated September 19, 1995
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) related to the replacement of
the steam generators at McGuire, Units 1 and 2. Currently, the steam
generators in place at the McGuire units are Westinghouse Model ``D''
type preheat steam generators. The tube degradation levels in the
generators has affected the reliability of the units. Therefore, these
generators are scheduled to be replaced with feedring steam generators
designed by Babcock and Wilcox International.
In the licensee's September 19, 1995, supplement, proposed changes
were made to TS Table 2.2-1, ``Reactor Trip System Instrumentation Trip
Setpoints,'' to change the programmed TAVG from 588.2 deg.F to
585.1 deg.F. This temperature was chosen based on returning the
secondary side steam pressure to the original value after replacement
of the steam generators. The licensee stated that 585.1 deg.F was the
assumed value for nominal full power TAVG in all applicable safety
analyses related to replacement of the steam generators.
The licensee also requested that the steam line safety valve lift
settings in Table 3.7-3, which was requested in the September 30, 1994,
application, be withdrawn. The licensee determined that these changes
are no longer needed.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards
[[Page 56367]]
consideration, which is presented below:
Operation of McGuire Nuclear Station in accordance with the
proposed changes to the Technical Specifications will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The low-low steam generator water
level reactor trip setpoint, the high-high steam generator water
level setpoint for turbine trip and feedwater isolation, and the
low-low steam generator water level setpoint for auxiliary feedwater
initiation are changing to support operation with the replacement
steam generators. These setpoints were chosen both to optimize plant
operation, and ensure that all applicable acceptance criteria are
met for licensing basis safety analysis. These setpoints do not
contribute to the initiation of any accident evaluated in the
McGuire FSAR [Final Safety Analysis Report] and have no adverse
impact on system operation, therefore it can be concluded that these
changes will not significantly increase the probability or
consequences of an accident evaluated in the FSAR.
The reduction in the primary to secondary leakage rate for
McGuire will not increase the probability of an accident evaluated
in the FSAR. This lower limit will require corrective action more
quickly than is currently required in the event that there is a
steam generator tube leak. This change will not significantly affect
the consequences of an accident previously evaluated. The allowable
leakage is being lowered because this leakage has a major impact on
the results of the offsite dose calculation for the locked rotor,
single uncontrolled rod withdrawal, and rod ejection events. The
taller tube bundle in the replacement steam generators will
potentially result in a longer period of tube bundle uncovery during
the above transients. The revised allowable leakages of 0.27 gpm
through all steam generators and 135 gallons per day through any one
generator ensure that the dose analysis results are within the
applicable fraction 10 CFR 100 limits.
The increase in Reactor Coolant System volume due to the
replacement steam generators will not increase the probability or
consequences of an accident previously evaluated. The increase in
volume has no effect on the probability of occurrence of any
accident evaluated in the FSAR. The mass and energy release due to
postulated loss of coolant accidents inside containment has been
analyzed to ensure that the peak containment pressure limit is not
exceeded. All Chapter 15 reanalysis which was required due to the
replacement steam generators assumed the new Reactor Coolant System
volume. Since the results of these analyses show the applicable
acceptance criteria continue to be met, it can be concluded that the
consequences of an accident previously evaluated are not
significantly increased due to this change.
* * * *
Operation of McGuire Nuclear Station in accordance with the
proposed changes to the Technical Specification will not create the
possibility of a new or different accident from any accident
previously evaluated. The proposed changes to revise the low-low
steam generator water level reactor trip setpoint, high-high steam
generator water level setpoint for turbine trip and feedwater
isolation, and low-low steam generator water level setpoint for
auxiliary feedwater initiation ensure that the appropriate
acceptance criteria for FSAR Chapter 15 transients which rely on
these functions are met for operation with the replacement steam
generators. The proposed change to lower primary to secondary
leakage for operation with the replacement steam generators will
require that corrective action be taken more quickly in the event
that steam generator tube leakage is experienced during operation.
As discussed in the technical justification, this will cause the
dose results for transients affected by tube bundle uncovery to be
within acceptable limits. .... The increase in Reactor Coolant
System volume is taken into account in the analysis of the mass and
energy release due to a postulated loss of coolant inside
containment and Chapter 15 events which have been reanalyzed due to
replacement of the steam generators. As discussed above, the
proposed changes will not introduce the possibility of a new or
different accident from any previously evaluated; they will ensure
that transients that take credit for these functions and dose
analyses meet applicable acceptance criteria for operation with the
replacement steam generators.
Operation of McGuire Nuclear Station in accordance with the
proposed changes to the Technical Specifications will not involve a
significant reduction in a margin of safety. The proposed changes
are being made to ensure that transients that rely on low-low steam
generator water level reactor trip setpoint, high-high steam
generator water level setpoint for turbine trip and feedwater
isolation, and low-low steam generator water level setpoint for
auxiliary feedwater actuation meet applicable acceptance criteria.
The reduction in allowable primary to secondary leak rate will
ensure that transients with dose analyses which are affected by the
replacement steam generators meet the current acceptable limits.
.... The proposed change in the Reactor Coolant System volume will
not involve a significant reduction in a margin of safety. The
increased volume affects the mass and energy release due to a
postulated loss of coolant accident inside containment and the other
Chapter 15 events which were reanalyzed due to replacement of the
steam generators. These events have been analyzed and the results
are within current acceptable limits. As discussed above, the
acceptance criteria for FSAR transients which are affected by these
proposed changes continue to be met, therefore there is no
significant reduction in the margin of safety.
Changes to the steam generator surveillance requirements will
simply delete inspection requirements and repair methods which are
no longer applicable after installation of the replacement steam
generators. The only exception to this is Surveillance Requirement
4.4.5.4.a.9. This requirement is modified to clarify that the
manufacturer will perform the hydrostatic test for the replacement
steam generators. This change will not affect the probability or
consequences of an accident previously evaluated, the purpose of the
preservice inspection is to establish the baseline condition of the
tubing. The baseline condition of the tubing in the replacement
steam generators will be established prior to installation. The
possibility of a new or different accident from any previously
evaluated will not be created. No new accident initiation mechanisms
will be introduced by this change, and the intent of the
requirement, to establish the baseline condition of the tubing, will
be met. Since the baseline condition of the tubing will be obtained
for use in the monitoring of tubing degradation, as is currently
required by the surveillance requirement, there will not be a
significant reduction in the margin of safety.
The changes to Technical Specification 6.9.1.9 are
administrative in nature. These changes are being made to reflect
the most recent revisions of DPC-NE-3002 and DPC-NE-3000, which
include changes associated with the replacement steam generators.
These topical reports revisions will be reviewed and approved for
use regarding Catawba and McGuire Nuclear Stations. Since these
changes are administrative in nature, no significant hazards
considerations are involved.
Proposed revision to TS Table 2.2-1, Reactor Trip System
Instrumentation Trip Setpoints:
proposed change to the Technical Specifications does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. Changing the value for TAVG
in Notes 1 and 2 of Table 2.2-1 will update the value to agree with
the TAVG assumed in the applicable safety analyses for
replacement of the steam generators. Acceptable results were
obtained for all required reanalyses. The probability of an accident
will not be significantly affected by operation with the new
TAVG value, because all equipment will be operated within
acceptable design limits. The consequences of previously evaluated
accidents which are affected by this change have been evaluated, and
have been determined to be within acceptable limits.
This proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated. This
change does not change the physical configuration of the plant, and
all analyses which are affected by replacement of the steam
generators have been determined to have acceptable results assuming
this value for TAVG.
This proposed change to the Technical Specifications will not
involve a significant reduction in the margin of safety. All safety
analyses which were affected by replacement of the steam generators
assumed this value for TAVG and the results were determined to
be within previously acceptable limits.
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.
[[Page 56368]]
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: September 4, 1993, as supplemented on
February 16, 1994, and August 4, 1995.
Description of amendment request: The proposed amendment would
revise the Arkansas Nuclear One Industrial Security Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The NRC staff's review is presented below.
The accident mitigation features of the plant are not affected by
the proposed compensatory measures for protecting the site during
periods when security systems are degraded and therefore no decrease
occurs in the effectiveness of the security program to protect against
radiological sabotage or increased risk to the public health and
safety. This is due to continued compliance with existing regulatory
requirements and other commitments within the security plan. These
changes have no impact on the design basis security threat and
accordingly do not create the possibility of a new or different kind of
accident. New systems, modes of equipment operation, failure modes or
other plan situations are not introduced by these changes. The proposed
changes allow flexibility for the use of compensatory measures and do
not change any safety limits, LCOs, or surveillance requirements on
equipment to operate the plant.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 24, 1995, as supplemented or
supercedes letters dated May 30, and June 20, 1995
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) on containment systems to
reflect the adoption of requirements of 10 CFR Part 50, Appendix J,
Option B, and implementation of a performance-based containment leak
rate testing program at River Bend Station. The licensee letters dated
May 20, and June 20, 1995, requested an exemption to Appendix J which
subsequently became Option B to the appendix. Those letters were
noticed in the Federal Register on July 5, 1995 (60 FR 35079).
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to
initiation of any accidents previously evaluated. Thus, the proposed
change cannot increase the probability of any accident previously
evaluated.
The proposed change potentially affects the leak-tight integrity
of the containment structure designed to mitigate the consequences
of a loss-of coolant accident (LOCA). The function of the
containment is to maintain functional integrity during and following
the peak transient pressures and temperatures which result from any
loss-of-coolant accident (LOCA)[LOCA]. The containment is designed
to limit fission product leakage following the design basis LOCA.
Because the proposed change does not alter the plant design, only
the frequency of measuring Type B and C leakage, the proposed change
does not directly result in an increase in containment leakage.
However, decreasing the test frequency can increase the probability
that a large increase in containment leakage could go undetected for
an extended period of time. Based upon the results of the periodic
containment Type A or Integrated Leak Rate Tests (ILRTs) and Type B
and C or Local Leak Rate Tests (LLRTs) surveillance tests, this is
not expected during the remaining life of the plant. The risk
resulting from the proposed changes is as follows:
Type A Testing
NUREG/CR-4330 (NRC86) found that the effect of containment
leakage on overall accident risk is small since risk is dominated by
accident sequences that result in failure or bypass of the
containment. It also determined that on an expected individual dose
basis, the effect of containment leakage is small.
Industry wide, ILRTs have only found a small fraction of the
leaks that exceed current acceptance criteria. Only three percent of
all leaks have a potential for remaining undetected for longer
periods of time. In addition, when leakage has been detected by
ILRTs, the leakage rate has been only about two times the allowable
leakage rate.
NUREG-1493 found that these observations, together with the
insensitivity of reactor accident risk to the containment leakage
rate, show that reducing the Type A leakage test frequency would
have a minimal impact on public risk.
Type B and C Testing
NUREG-1493 found that while Type B and C tests can identify the
vast majority (greater than 95 percent) of all potential leakage
paths, performance-based alternatives to current local leakage-
testing requirements are feasible without significant risk impacts.
The risk model used in NUREG-1493 suggests hat the number of
components tested would be reduced by about 60 percent with less
than a three-fold increase in the incremental risk due to
containment leakage. Since, under existing requirements, leakage
contributes less than 0.1 percent of overall accident risk, the
overall impact is very small. NUREG-1493 found that while the
extended testing intervals for Type B and C tests led to minor
increases in potential offsite [off-site] dose consequences, the
actual increase in on-site (worker) doses exceeded (by at least an
order of magnitude) the potential off-site dose consequences.
EPRI Research Project Report TR-104285, ``Risk Impact Assessment
of Revised Containment Leak Rate Testing Intervals,'' also concluded
that a relaxation of the test intervals for Type B and C
penetrations results in a negligible increase in total plant risk.
Based on the above EOI [Entergy Operation, Inc.] has concluded
that the proposed change will not result in a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. This change involves the reduction in
Type B and C test frequency. The methods of performing the tests are
not changed. No new accident modes are created by extending the
testing intervals. No safety-related equipment or safety functions
are altered as a result of this change. Extending
[[Page 56369]]
the test frequency has no influence on , nor does it contribute to, the
possibility of a new or different kind of accident or malfunction
from those previously analyzed.
3. The request does not involve a significant reduction in a
margin to safety.
The proposed change only affects the frequency of Type A, B, and
C testing and does not change the methodology for performance of the
testing. However, the proposed change can increase the probability
that a large increase in leakage could go undetected for an extended
period of time. Operational experience has shown that the leak
tightness of the containment has been maintained significantly below
the allowable leakage limit. In addition, NUREG-1493 has determined
that, under several different accident scenarios, the risk of
radioactivity release from containment is negligible with the
implementation of these proposed changes.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite [off-site] dose
consequences of postulated accidents which are directly related to
containment leakage rate. The containment isolation system is
designed to limit leakage to La which is defined by the RBS
Technical Specifications to be 0.26 percent by weight of the
containment air per 24 hours at 7.6 psig (Pa). The limitation
on containment leakage rate is designed to ensure that total leakage
volume will not exceed the value assumed in the accident analyses at
the peak accident pressure (Pa) or 7.6 psig. The margin to
safety for the offsite [off-site] dose consequences of postulated
accidents directly related to the containment leakage rate in
maintained by meeting the 1.0 La
No change in the method of testing is being proposed. The Type B
and C tests will continue to be done at full pressure (Pa) or
greater. Other programs are in place to ensure that proper
maintenance and repairs are performed during the service life of the
primary containment and systems and components penetrating the
primary containment.
As a result, EOI had concluded that the proposed change will not
result in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: September 28, 1995
Description of amendment request: The proposed change modifies
Technical Specification 3/4.8.1.2, ``Electrical Power Sources -
Shutdown.'' The surveillance requirement 4.8.1.2 is clarified by a Note
to identify those surveillances which are required to be performed
during Modes 5 and 6.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
No component modification, system realignment, or change in
operations will occur which could affect the probability of any
accident or transient. The proposed addition of a Note will provide
guidance on which surveillances are required to be performed in
Modes 5 and 6. The Note will preclude rendering operable DGs
inoperable, and/or preclude de-energizing a required ESF bus or
disconnecting a required offsite circuit during the performance of
the surveillance requirement. Proposed changes do not eliminate any
testing requirements, they simply clarify which tests will be
performed in Modes 5 and 6, and which are required to be performed
prior to entry into Mode 4. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously analyzed.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
No component modification, system realignment, or change in
operating procedure is required to implement the proposed change.
The proposed change reduces the possibility of a single event
impacting the operability of an ESF bus or its DG simultaneously.
Therefore, these changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The proposed change will not alter any assumptions, initial
conditions, or results of any accident analyses. The Class 1E
equipment assumed available in the accident analyses and their
designed capability to mitigate the consequences of any postulated
accidents will not be changed. The addition of a Note to clarify the
surveillance requirements will not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: September 28, 1995
Description of amendment request: The proposed changes relocate
``Reactor Coolant System - Chemistry'' Technical Specification 3/4.4.7
(Salem Unit 1) and 3/4.4.8 (Salem Unit 2) and their associated Bases to
the Salem Updated Final Safety Analysis Report (UFSAR) and the
Surveillance Requirements and Limiting Conditions for Operation to
applicable plant procedures controlled by the 10 CFR 50.59 process.
Also, the applicability will be changed from ``At all times'' to
``Modes 1, 2, 3, 4, 5 and 6.''
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. Specifically, changing the Applicability from
``At all times'' to ``Modes 1, 2, 3, 4, 5 and 6'' by this submittal
will not alter established chemistry for chlorides, fluorides and
dissolved oxygen of the Reactor Coolant System. The relocation of
this Surveillance Requirement/LCOs and Bases to plant procedures and
the UFSAR respectively, will continue to ensure that the chemistry
analysis of the Reactor Coolant System water is monitored and
controlled. Changing the Applicability from ``At all times'' to
``Modes 1,2,3,4,5 and 6'' represent changes that do not affect plant
safety and do not alter existing accident analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes are procedural in nature concerning the
location of the descriptive information and surveillance
requirements for Reactor Coolant System Chemistry. Removing these
specifications from the Technical Specifications and
[[Page 56370]]
placing them in the UFSAR and plant procedures will not alter the
maintenance of the Reactor Coolant System Chemistry or the ability
to monitor its intended functions. Therefore, these changes will not
create a new or unevaluated accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
The proposed changes relocate the Reactor Coolant System
Chemistry Requirements/LCOs from the Technical Specifications to the
UFSAR and plant procedures in accordance with guidance provided by
the NRC Final Policy Statement (58 FR 39132) regarding the
improvement of Technical Specifications. The requirements that will
reside in the UFSAR and plant procedures for the Reactor Coolant
System Chemistry will ensure that the ability to determine chloride,
fluoride and dissolved oxygen concentrations in the Reactor Coolant
System is properly maintained and that the maintenance of the
Reactor Coolant System Chemistry will be commensurate with its
safety significance. Therefore, the proposed changes will not
involve a significant reduction in any margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
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 26, 1995
Description of amendments request: The amendments would revise
Technical Specification (TS) Section 4.6.1.3 to incorporate
improvements to containment air lock testing referenced in Chapter 3.6,
``Containment Systems,'' of NUREG-1431, ``Standard Technical
Specifications, WestinghousePlants.''
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 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 does not involve any change to the
configuration or method of operation of any plant equipment used to
mitigate the consequences of an accident. Containment leakage is an
assumption in the safety analysis of the loss of coolant accident
and the rod ejection accident. Changes to the containment air lock
door seal test acceptance criteria will have no impact on the
radiological consequences of these accidents since the plant safety
analysis is based on the assumption that the containment leaks at
its design leak rate of 0.15 percent per day for the first 24 hours
and 0.075 percent per day thereafter for each of these accidents.
The change to the surveillance requirement meets the intent of the
guidance in NUREG-1431. Primary containment integrity ensures that
the release of radioactive materials from the containment atmosphere
will be restricted to those leakage paths and associated leak rates
assumed in the accident analysis. The limitations on closure and
leak rate for the containment air locks are required to meet these
restrictions on containment integrity. These changes do not increase
the probability that the 10 CFR [Part] 100 limits will be exceeded.
The change to the surveillance requirement does not impose any new
safety analyses limits or alter the plants ability to detect and
mitigate events. Therefore, 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 involves a revision to the Technical
Specifications to meet the intent of the guidance of NUREG-1431, and
does not necessitate a physical alteration of the plant or change in
parameters governing normal plant operation. The change has not
effect on the plant's compliance with the requirements of Appendix
J. The revision of the acceptance criteria for the air lock door
seal test will improve the FNP [Farley Nuclear Plant] current
testing criteria while maintaining an acceptable level of safety.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety. The revision of the acceptance criteria of
the air lock door seal test will decrease the overall test burden
without decreasing the margin of safety. The overall leakage rate of
the air lock continues as less than or equal to 0.05La and the
plant safety analysis continues to be based ont he assumption that
the containment leaks at its design leak rate of 0.15 percent per
day for the first 24 hours and 0.075 percent per day thereafter for
each of these accidents. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: July 28, 1995
Description of amendment request: The proposed amendment would
clarify the limiting condition for operation for TS 3.8.1.1 and 3.8.1.2
from ``independent'' circuit to ``qualified'' circuit; explain in the
Bases the requirements for operability of an offsite circuit; delete
the STAGGERED TEST BASIS scheduling requirement to perform emergency
diesel generatorsurveillances; explain in the Bases an acceptable
method for verification of Emergency Diesel Generator speed for
surveillance requirements (SR) 4.8.1.1.2.a.4 and 4.8.1.1.2.c.4; remove
a surveillance test extension that has expired for SR 4.8.1.1.1.b; add
an exception for SR 4.8.1.1.2.c.5 and 4.8.1.1.2.c.7 to SR 4.8.1.2; and
revise Bases 3.0.5 to reflect the clarification from ``independent''
circuit to ``qualified'' circuit.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation ofthe Davis-Besse Nuclear Power Station, Unit No. 1 in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed changes do not
make a change to any accident initiator, initiating condition or
assumption. The proposed changes do not involve a significant change
to the plant design or operation. The proposed changes do not affect
the safety function of the offsite circuits or the emergency diesel
generators (EDGs).
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate assumptions used in evaluating the radiological
consequences of
[[Page 56371]]
an accident, do not alter the source term or containment isolation and
do not provide a new radiation release path or alter potential
radiological releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not introduce a new or different accident initiator or
introduce a new or different equipment failure mode or mechanism.
3. Not involve a significant reduction in a margin of safety
because the proposed changes do not reduce the margin to safety
which exists in the present Technical Specifications [TS] or Updated
Safety Analysis Report. The operability requirements of the TS are
consistent with the initial condition assumptions of the safety
analyses. Further, the proposed changes do not affect the Action
statement requirements for the various levels of degradation in the
offsite [power] circuits or EDGs.
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: September 29, 1995
Description of amendment request: The proposed amendment would
increase the minimum available borated water volume requirement for the
boric acid addition system, the minimum and maximum boron concentration
requirements for the borated water storage tank, the minimum boron
concentration requirement for the core flood tanks; modify the
surveillance requirements for trisodium phosphate dodecahydrate; and
modify the refueling boron concentration and the associated Action
statement. These proposed changes will affect the following Technical
Specification sections: 3/4.1.2.8, Reactivity Control Systems - Borated
Water Sources - Shutdown; 3/4.1.2.9, Reactivity Control Systems -
Operating; 3/4.5.1, Emergency Core Cooling Systems (ECCS) - Core
Flooding Tanks; 3/4.5.2, Emergency Core Cooling Systems - ECCS
Subsystems - Tavg [plus or minus] 280 deg.F; 3/4.5.4, ECCS - Borated
Water Storage Tank; 3/4.9.1, Refueling Operations - Boron
Concentration; Bases 3/4.1.2, Boration Systems; Bases 3/4.5.2 and 3/
4.5.3, ECCS Subsystems; and Bases 3/4.9.1 Boron Concentration.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions, or assumptions are significantly affected by the
proposed changes.
The proposed changes to the Technical Specifications and their
Bases increase the minimum volume of the Boric Acid Addition System
(BAAS), the minimum boron concentration of the Borated Water Storage
Tank (BWST) and Core Flooding Tanks (CFTs), the maximum boron
concentration of the BWST, and the minimum volume of trisodium
phosphate dodecahydrate (TSP) in Containment (CTMT). Administrative
changes to these Technical Specifications have also been proposed.
These changes ensure adequate boration capability is maintained for
normal operations, that adequate Shutdown Margin (SDM) can be
achieved following an accident, and that the assumed post-Loss of
Coolant Accident (LOCA) pH can be achieved. Therefore, as stated
above, these proposed changes do not significantly affect accident
initiators, conditions, or assumptions.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
change the source term, CTMT isolation, or allowable releases.
In particular, maintaining the appropriate amount of TSP will
ensure the assumed pH will be achieved, the assumption of source
term with respect to iodine retention will be maintained, and the
radiological consequences of a previously evaluated accident will
not be increased.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes.
These changes ensure that the assumptions used for initial and
final conditions of SDM, pH, and source term are maintained. Also,
the Environmental Qualification (EQ) and seismic requirements have
been verified to be adequate to maintain the adequacy of Structures,
Systems, and Components (SSCs) during assumed accident conditions.
3. Not involve a significant reduction in a margin of safety
because the proposed changes to the minimum volume and boron
concentration for the BAAS, BWST, and CFTs ensure that the margin of
safety for reactor subcriticality is maintained at all times for
future longer fuel cycles, including the upcoming Cycle 11.
The proposed increase in the BWST maximum boron concentration is
set at the conservative limit for post-LOCA boron precipitation
concerns. Therefore, the existing margin of safety with respect to
post-LOCA boron precipitation is maintained.
The proposed increase in the minimum TSP volume requirement
maintains the same margin of safety with respect to post-LOCA pH,
time for dissolution, iodine retention, and chloride stress
corrosion of austenitic stainless steels. The TSP capacity margin of
approximately 40 cubic feet included in the minimum TSP volume
requirement will not result in increasing the pH above the
previously approved pH limit of 11. This reserve capacity adds
margin to ensure adequate minimum pH is achieved.
The proposed removal of the 1800 ppm refueling boron
concentration requirement does not reduce the margin of safety
because the requirement of maintaining keff [less than or equal to]
0.95 is alone sufficient to ensure that the accident analysis
assumptions are satisfied.
The proposed change to the boration rate requirement of the
LCO 3.9.1 Action statement does not reduce the margin of safety
because the proposed boration rate of 12 gpm of 7875 ppm boric acid
solution is equivalent to the present boration rate of
10 gpm of 8750 ppm boric acid solution.
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: October 2, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 5.0, ``Design Features,''
by adding a site location description, remove site area
[[Page 56372]]
maps, remove containment and reactor coolant system design parameters,
remove the description of the meteorological tower location, remove
component cyclic or transient limits, and revise the fuel assembly
description to include the use of ZIRLO clad fuel rods.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station Unit Number 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions or assumptions are affected by the proposed changes to
Section 5.0, Design Features, of the Technical Specifications. These
changes are proposed to add a site location description, remove site
area maps, remove containment and reactor coolant system design
parameters, remove the description of the meteorological tower
location, remove component cyclic or transient limits, and revise
the fuel assembly description to include the use of ZIRLO clad fuel
rods.
Under the proposed changes, Technical Specifications (TS)
Section 5.0 would continue to satisfy the applicable requirements of
Section 182.a of the Atomic energy Act of 1954, and 10 CFR
50.36(c)(4). Further, the proposed changes are consistent with
NUREG-1430, ``Standard Technical Specifications for Babcock and
Wilcox Plants,'' Revision 1. The information proposed for removal
from existing TS 5.0 is presently included in the Updated Safety
Analysis Report (USAR) or is being proposed to be added to the USAR,
hence sufficient controls exist under 10 CFR 50.59 to ensure that
future changes to these items are acceptable.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected by the proposed changes. As described
above, these changes are consistent with the ``Standard Technical
specifications for Babcock and Wilcox Plants'' (NUREG-1430) and are
administrative changes. The proposed changes do not alter the source
term, containment isolation, or allowable releases. The proposed
changes, therefore, will not increase the radiological consequences
of a previously evaluated accident.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes, which involve only administrative controls. As described
above, these changes are consistent with the ``Standard Technical
Specifications for Babcock and Wilcox Plants'' (NUREG-1430) and are
administrative changes. The proposed changes do not alter any
accident scenarios.
3. Not involve a significant reduction in a margin of safety
because the proposed changes are administrative and do not reduce or
adversely affect the capabilities of any plant structure, systems or
components.
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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: September 6, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.3.1 to reflect a change in the
maximum initial enrichment for reload fuel. The amendment would also
change the maximum reference Kinfinity for storage in Region 1 of
the spent fuel pool and TS Figure 3.9-1 to reflect a change in the
maximum initial enrichment for storage in Region 2.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 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.
An increase to a maximum initial enrichment of 5.0 w/o U-235
does not involve an increase in the probability or consequence of an
accident or other adverse condition over previous evaluations.
Because of the conservative techniques and assumptions used to
evaluate the maximum possible neutron multiplication factor, there
is reasonable assurance that criticality safety is maintained when
storing fuel assemblies of up to and including 5.0 w/o U-235 in the
spent fuel storage racks under both normal and postulated accident
conditions. For example, the calculations for non-accident
conditions ignore the 2000 ppm soluble boron in the spent fuel pool
calculations, thus resulting in conservative values of the
multiplication factor. Storing fuel in the Region 1 configuration
which meets the IFBA [integral fuel burnable absorber] versus
enrichment curve (Figure 3 of Attachment 6) results in a maximum
multiplication factor of 0.9481, including all biases and
uncertainties.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
An increase to a maximum initial enrichment level of 5.0 w/o U-
235 does not create the possibility of a new or different kind of
accident or condition over previous evaluations. An increase to the
enrichment level of 5.0 w/o U-235 involved performing extensive
evaluations to develop the IFBA versus enrichment curve for V-5
fuel. Use of dual code packages ensures that the spent fuel pool
Region 1 criticality limits are not exceeded.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
An increase in the maximum initial enrichment level to 5.0 w/o
U-235 does not involve a reduction in the margin of safety. As
discussed above, in all cases the multiplication factors for worst
case assumptions fall considerably below the criticality limits and
do not represent any reductions in margin. An increase to the
initial enrichment level of 5.0 w/o U-235 does not adversely impact
operation of the various plant systems, i.e. HVAC [heating,
ventilation, and air conditioning], spent fuel pool cooling, or
radiological control systems.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: William H. Bateman
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 6, 1995
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
4.2.b, ``Steam Generator Tubes,'' its associated bases, and Figure TS
4.2-1 by redefining the pressure boundary for Westinghouse mechanical
hybrid expansion joint (HEJ) steam generator (SG) tube sleeves.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the
[[Page 56373]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
The proposed change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist.
1. Operation of the KNPP in accordance with the proposed license
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Mechanical testing has shown that the inherent structural
strength of the HEJ joint provides sufficient integrity such that
the tube rupture capability recommendations of RG [Regulatory Guide]
1.121 are met, even for instances of 100 percent throughwall,
360 deg. circumferentially oriented degradation in the HEJ HRLT
[hardroll lower transition] region. Structural integrity
recommendations consistent with RG 1.121 are supplied
for all tube degradation 1.1 inch or greater below the bottom of
the HEJ HRUT [hardroll upper transition]. Based on test data, a
bounding SLB [steam line break] leak rate of 0.033 gpm for
indications between 1.1 and 1.3 inch below the bottom of the HRUT is
applied. As the leakage data base is expanded and statistical basis
established, this SLB leakage allowance may be reduced. For
indications existing greater than 1.3 inch below the bottom of the
HRUT, SLB event leakage can be neglected.
Additional prevention from tube rupture is inherently provided
by the HEJ geometry. For RCS [reactor coolant system] release rates
to exceed the normal makeup capacity of the plant, the tube must be
postulated to experience a complete circumferential separation at
the lower transition, and become axially displaced by 3 to 3.25
inches, resulting in complete geometric disassociation between the
tube and sleeve resulting in sufficient flow area to support leakage
in excess of makeup capacity. During the 1989 plug top release event
at North Anna Unit 1, primary to secondary release rates were
calculated to be less than 80 gpm, for a flow area approximately
four times larger than the flow area created by a tube which was
axially displaced by about 1.25 to 1.5 inch. Analysis of the steam
generator indicates that at a 95 percent cumulative probability, the
tube would experience an axial displacement of less than the 1.1
inch boundary. At this level of axial displacement, a ring of metal
to metal contact would remain between the tube and sleeve, and
leakage would be far less than makeup. Projected leakage at this
point is expected to be less than 2.5 gpm. Therefore, implementation
of the proposed repair boundary will not result in tube rupture,
even for a tube postulated to not behave as predicted by the
available test and pulled tube data.
The proposed technical specification change to support the
implementation of the HEJ sleeve tube pressure boundary for parent
tube degradation in the HEJ HRLT region does not adversely impact
any other previously evaluated design basis accident or the results
of accident analyses for the current technical specification minimum
reactor coolant system flow rate. Plugging limit criteria are
established using the guidance of RG 1.121. Furthermore, per RG 1.83
recommendations, the sleeved tube assembly can be monitored through
periodic inspections with present eddy current techniques.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Implementation of the revised pressure boundary will not
introduce significant or adverse changes to the plant design basis.
Mechanical testing of degraded sleeve joints supports the
conclusions of the calculations that the sleeve retains structural
(tube burst) capability consistent with RG 1.121. As with initial
installation of sleeves, implementation of the relocated pressure
boundary cannot interact with other portions of the RCS. Any
hypothetical accident as a result of potential tube degradation in
the HEJ HRLT region of the tube is bounded by the existing tube
rupture accident analysis. Neither the sleeve design nor
implementation of the tube repair boundary defined on Figure TS 4.2-
1 affects any other component or location of the tube outside of the
immediate area repaired.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
The safety factors used in the establishment of the HEJ sleeved
tube pressure boundary are consistent with the safety factors in the
ASME [American Society of Mechanical Engineers] Boiler and Pressure
Vessel Code used in steam generator design. Based on the sleeved
tube geometry, it is unrealistic to consider that application of the
revised pressure boundary could result in single tube leak rates
exceeding the normal makeup capacity during normal operating
conditions. The pressure boundary established ... has been developed
using the methodology of RG 1.121. The performance characteristics
of postulated degraded parent tubes of HEJ tube/sleeve joints have
been verified by testing to retain structural integrity and preclude
significant leakage during normal and postulated accident
conditions. Testing indicates that postulated circumferentially
separated tubes which the repair boundary addresses would not
experience axial displacement during either normal operation or SLB
conditions. The existing offsite dose evaluation performed for KNPP
in support of the voltage based plugging criteria for axial ODSCC
[outside diameter stress corrosion cracking] at TSP [tube support
plate] intersections established a faulted loop primary to secondary
leak rate of 34.0 gpm using technical specification dose equivalent
Iodine-131 activity levels. Following implementation of the
criteria, postulated leakage from all sources must not exceed 34.0
gpm in the faulted loop. Maintenance of this limit will ensure that
offsite doses would not exceed the currently accepted limit of a
small fraction of the 10 CFR 100 guidelines. The repair boundary
uses a conservatively established ``per indication'' leak rate for
estimation of SLB leakage. This leak rate is applied to all
indications left in service as a result of the tube repair boundary,
including non-throughwall indications and a limited number of
indications of circumferential throughwall extent.
For a postulated indication whose performance is not
characteristic of the test and pulled tube data, and which would
experience axial displacement at the 95 percent cumulative
probability value following a postulated SLB event with no operator
intervention, leakage would not be expected to result in an
uncontrolled release of reactor coolant in excess of normal makeup
capacity.
For the three removed tube sleeve samples and nearly 1,000 PTIs
[parent tube indications] detected in the field, there were no
instances of degradation of elevations (multiple expansion
transitions) on either side of the hardroll expansion in the same
tube. This includes no instances on non-detected degradation in the
upper hydraulic and hardroll upper expansion transitions for the
removed tubes. One tube was identified in the most recent KNPP
inspection with two separate circumferential crack elevations within
the HRLT. Rapidly occurring degradation would not be expected at the
upper transitions, based partly on the field inspection results. The
available inspection results include two inspection programs (1994
and 1995) at Kewaunee and one at Point Beach Unit 2 (1994). Through
these three inspection programs, approximately 11,000 HEJ sleeved
tubes have been inspected using advanced ET [eddy current testing]
techniques.
The portions of the installed sleeve assembly which represent
the reactor coolant pressure boundary can be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the requirements of Regulatory Guide 1.83.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments
[[Page 56374]]
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.
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: October 6, 1995
Description of amendment request: Revise the Technical
Specifications to change the definition of the F* distance.
Date of publication of individual notice in Federal Register:
October 16, 1995 (60 FR 53648)
Expiration date of individual notice notice: November 15, 1995
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendment: September 13, 1995, as
supplemented by letter dated October 19, 1995
Brief description of amendment request: The proposed amendments
would revise Technical Specification (TS) Section 15.1,
``Definitions,'' the basis for TS Section 15.3.1.G, ``Operational
Limitations,'' and TS Figure 15.2.1-2, ``Reactor Core Safety Limits,
Point Beach Unit 2.'' The proposed changes would reduce the reactor
coolant system raw measured total flow rate limit and reflect new
reactor core safety limits for Unit 2.Date of individual notice in
Federal Register: October 24, 1995 (60 FR 54527)
Expiration date of individual notice notice: November 8, 1995
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, 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.
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 6, 1995
Brief description of amendments: The amendments extend the nominal
surveillance interval requirements of selected safety systems
instruments form 18 months to a refueling interval of 24 months.
Date of issuance: October 19, 1995
Effective date: As of the date of issuance to be implemented within
30 days for Unit 2 and prior to restart of the spring 1996 refueling
outage for Unit 1.
Amendment Nos.: 208 and 186
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35061) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated October 19, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library,
Prince Frederick, Maryland 20678
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois Docket Nos. 50-10, 50-237 and 50-249, Dresden
Nuclear Power Station, Units 1, 2 and 3, Grundy County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and
2, LaSalle County, Illinois Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station,
Units 1 and 2, Lake County, Illinois
Date of application for amendments: April 24, 1995, as supplemented
August 1 and September 14, 1995.
Brief description of amendments: The amendments would relocate the
requirements for the ``Review, Investigative and Audit Functions'' and
frequencies of the quality assurance (QA) program from the
administrative controls section of the TS to the appropriate sections
of the licensee's Quality Assurance Topical Report (QATR), CE-1-A,
Revision 65. In addition, the proposed TS changes include title changes
to reflect the reorganization of the licensee's Nuclear Operations
Division and miscellaneous administrative and editorial changes.
Date of issuance: October 20, 1995
Effective date: October 20, 1995
Amendment Nos.: 75, 75, 67, 67, 38, 141, 135, 107, 93, 163, 159,
171, and 158
Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77,
DPR-2, DPR-19, DPR-25, NPF-11 NPF-18, DPR-29, DPR-30, DPR-39 and DPR-
48: The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45175) and September 20, 1995 (60 FR 48726). The August 1 and September
14, 1995, letters provided clarifying information that did not change
the initial proposed no significant hazards consideration determination
or expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 20, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481; for Dresden, Morris Area Public Library
District, 604
[[Page 56375]]
Liberty Street, Morris, Illinois 60450; for LaSalle, Jacobs Memorial
Library, Illinois Valley Community College, Oglesby, Illinois 61348;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021; and for Zion, Waukegan Public Library, 128 N. County
Street, Waukegan, Illinois 60085
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of application for amendments: June 30, 1995
Brief description of amendments: The amendments modify the
surveillance requirements for the emergency diesel generators.
Date of issuance: October 16, 1995
Effective date: October 16, 1995
Amendment Nos.: 170 and 157
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47615) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 16, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: April 4, 1995
Brief description of amendment: The amendment deletes requirements
associated with part length control element assemblies.
Date of issuance: October 12, 1995
Effective date: October 12, 1995
Amendment No.: 169
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37090) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 12, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: March 17, 1995
Brief description of amendment: The amendment deletes requirements
associated with surveillance to verify position stops for High Pressure
Safety Injection Emergency Core Cooling System throttle valves.
Date of issuance: October 18, 1995
Effective date: October 18, 1995
Amendment No.: 170
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37089) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 18, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: April 4, 1995, as supplemented by letter
dated October 12, 1995
Brief description of amendment: The amendment revises the
containment cooling response time to reduce the likelihood of a water
hammer event in service water piping.
Date of issuance: October 26, 1995
Effective date: October 26, 1995
Amendment No.: 171
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 1995 (60 FR
37090) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 26, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: February 28, 1994
Brief description of amendments: The amendments delete the minimum
frequency criteria prescribed for quality assurance audits from
Administrative Controls sections 6.5.2.8 and 6.8.4 of the Technical
Specifications (TS). Audit periodicity will thereby be controlled by
the program described in the Florida Power and Light Company (FPL)
Topical Quality Assurance Report.
Date of issuance: October 25, 1995
Effective date: October 25, 1995
Amendment Nos.: 140 and 80
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17599) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 25, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: July 26, 1995
Brief description of amendments: These amendments consist of
administrative corrections and clarifications.
Date of issuance: October 17, 1995
Effective date: October 17, 1995
Amendment Nos. 177 and 171Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47619) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 17, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: July 26, 1995
Brief description of amendments: These amendments consist of
administrative corrections and clarifications.
Date of issuance: October 17, 1995
Effective date: October 17, 1995
Amendment Nos.: 178 and 172Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47619) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 17, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199
[[Page 56376]]
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: February 13, 1995, as
supplemented April 21, 1995, and August 7, 1995.
Brief description of amendment: The proposed amendment deletes the
audit requirements from the Duane Arnold Energy Center Technical
Specifications (TS) and adds them to the Quality Assurance Program.
Date of issuance: October 17, 1995
Effective date: October 17, 1995
Amendment No.: 213
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16190) The additional information contained in the supplemental letters
dated April 21, 1995, and August 7, 1995, was clarifying in nature and
did not change the NRC staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 17, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: March 31, 1995
Brief description of amendments: The amendments revise Technical
Specification (TS) surveillance requirements for safety-related pump
testing to eliminate recirculation alignments. In addition, specific
test parameters, discharge pressures, and flows associated with these
pumps are removed from the TS and will be controlled by the Inservice
Testing Program.
Date of issuance: October 17, 1995
Effective date: October 17, 1995, with full implementation within
45 days
Amendment Nos.: 203 and 188
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 21, 1995 (60 FR
32368) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 17, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: February 1, 1995
Brief description of amendment: The amendment revises Technical
Specification 3.6.13 and associated Bases to permit the controls and
instruments from both Remote Shutdown Panels to be considered when
assuring that one complete set of controls and instruments is operable.
The changes also allow 30 days to restore an inoperable function to
operable status, remove MODE 3 (hot shutdown) from the existing
requirement for operability, and revise the LIMITING CONDITION FOR
OPERATION ACTION to require achieving hot shutdown in 12 hours instead
of cold shutdown in 36 hours. An additional change permits the operator
30 days to establish an alternate method of monitoring a parameter (and
90 days to restore the function) when the function is inoperable.
Date of issuance: October 16, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 155
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11135) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 16, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: March 29, 1995
Brief description of amendment: The amendment modifies the current
Technical Specifications that have cycle-specific parameter limits in
the Core Operating Limits Report to include an additional cycle-
specific parameter and its supporting methodologies.
Date of issuance: October 18, 1995
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 120
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24912) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 18, 1995.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
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit
No. 3, York County, Pennsylvania
Date of application for amendment: September 1, 1995
Brief description of amendment: The amendment deleted License
Condition 2.C.(5) which restricts power levels to no less than seventy
percent in the coastdown condition.
Date of issuance: October 17, 1995
Effective date: As of date of issuance
Amendment No.: 215
Facility Operating License No. (DPR-56): This amendment revised the
Facility Operating License. Public comments requested as to proposed no
significant hazards consideration: Yes. (60 FR 48530). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by October 18, 1995, but indicated that if the Commission makes
a final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final no significant hazards consideration
determination are contained in a Safety Evaluation dated October 17,
1995.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. Vice
President and General Counsel, PECO Energy Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
[[Page 56377]]
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: May 19, 1995
Brief description of amendments: The amendments revise the
Technical Specifications Table 3.3.3-3, ``Emergency Core Cooling System
Response Times'' to reflect the value of 60 seconds for the High
Pressure Coolant Injection system response time instead of 30 seconds
as previously specified.
Date of issuance: October 16, 1995
Effective date: For both units, as of the date of issuance and to
be implemented within 30 days.
Amendment Nos.: 102 and 66
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 5, 1995 (60 FR
35084) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 16, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: July 21, 1995
Brief description of amendment: The amendment revises TS Section
6.0 (Administrative Controls) to replace the title-specific list of
members on the Plant Operating Review Committee with a more general
statement of membership requirements, and expands the scope of
disciplines represented on the committee to include Nuclear Licensing
and Quality Assurance. The amendment also changes the following
management position titles: ``First Executive Vice President and Chief
Nuclear Officer'' to ``Chief Nuclear Officer'', ``Resident Manager'' to
``Site Executive Officer'', ``Shift Supervisor'' to ``Shift Manager'',
and ``Assistant Shift Supervisor'' to ``Control Room Supervisor.''
These changes in title do not affect the reporting relationships,
authority, or responsibilities of these positions. Finally, the
amendment also makes editorial corrections to the TSs.
Date of issuance: October 13, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 228
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47624) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 13, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: April 12, 1995.
Brief description of amendment: The amendment extends the
surveillance test intervals for the nuclear steam supply system to
support 24-month operating cycles. Surveillance test interval
extensions that are justified will be denoted as being performed
``every 24 months'' or ``at least once per 24 months'' consistent with
the guidance provided in Reference 1. Other surveillances currently
performed ``once each operating cycle,'' ``at least once during each
operating cycle,'' ``each refueling,'' or similar notation, that are
not being extended at this time will be denoted as being performed ``at
least once per 18 months.'' The NRC staff has determined that the
proposed TS changes follow the guidance of Generic Letter 91-04, and
are therefore acceptable.
Date of issuance: October 13, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 229
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24916) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 13, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: April 18, 1995
Brief description of amendment: This amendment changes Technical
Specification Table 4.3.7.1-1, ``Radiation Monitoring Instrumentation
Surveillance Requirements,'' to increase the channel functional test
interval from monthly to quarterly for each instrument.
Date of issuance: October 16, 1995
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 83
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42607) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 16, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: May 4, 1995
Brief description of amendment: This amendment changes Technical
Specification (TS) 3/4.6.1.8, ``Drywell and Suppression Chamber Purge
System,'' increasing the annual operational limit for the drywell and
suppression chamber purge system from 120 to 500 hours.
Date of issuance: October 16, 1995
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 84
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 16, 1995 (60 FR
42607) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 16, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
[[Page 56378]]
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of applications for amendment: November 30, 1994 and March 30,
1995, as supplemented by letter dated September 5, 1995.
Brief description of amendment: The change to TS Table 3.3.1-2,
``Reactor Protection System Response Times,'' TS Table 3.3.2-3,
``Isolation System Instrumentation Response Time,'' TS Table 3.3.3-3,
``Emergency Core Cooling System Response Times,'' and associated Bases,
eliminates the requirement to perform response time testing for certain
classes of equipment and transfers the requirements of the above-
referenced TS Tables to the Updated Final Safety Analysis Report.
Date of issuance: October 24,1995
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 85
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16198 and August 16, 1995 (60 FR 42606) The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
October 24, 1995. No significant hazards consideration comments
received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: May 20, 1994, as supplemented on
March 29, 1995
Brief description of amendment: The amendment revises Technical
Specifications to implement the NRC's Final Policy Statement on
Technical Specification Improvements for Nuclear Power Reactor by
relocating specifications that do not meet policy statement criteria to
the Final Safety Analysis Report.
Date of issuance: October 20, 1995
Effective date: Immediately, to be implemented within 120 days.
Amendment No.: 103
Facility Operating License No. NPF-30. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45036). The March 29, 1995, letter provided supplemental information
that did not change the initial proposed no significant hazards
consideration determination or expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 20, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: January 26, 1994, as
supplemented by letters dated December 1, 1994, and June 23, 1995
Brief description of amendments: These amendments revise Technical
Specification (TS) Section 15.3.0, ``General Considerations.'' This
section specifies the actions to be taken for conditions not directly
addressed in the action statements fo the TSs. In addition, changes to
the applicable bases (including the bases for TS 15.3.3) and editorial
changes are also included.
Date of issuance: October 12, 1995
Effective date: October 12, 1995
Amendment Nos.: 163 and 167
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12373) The December 1, 1994 and June 23, 1995, submittals provided
supplemental information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 12, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: April 17, 1995
Brief description of amendments: These amendments change TS
Sections 15.6.2, ``Organization,'' and 15.6.3, ``Facility Staff
Qualifications.'' The requirement for the Operations Manager to hold an
NRC Senior Reactor Operator's (SRO) license has been changed to provide
additional staffing flexibility.
Date of issuance: October 12, 1995
Effective date: October 12, 1995
Amendment Nos.: 164 and 168
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27346). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 12, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Dated at Rockville, Maryland, this 1st day of November 1995.
For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV,Office of Nuclear
Reactor Regulation
[Doc. 95-27543 Filed 11-7-95; 8:45 am]
BILLING CODE 7590-01-F