[Federal Register Volume 59, Number 3 (Wednesday, January 5, 1994)]
[Notices]
[Pages 615-639]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10105]
[[Page Unknown]]
[Federal Register: January 5, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice Applications and Amendments to Facility Operating
Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 13, 1993, through December 22,
1993. The last biweekly notice was published on December 22, 1993 (58
FR 67840).
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 P-223, Phillips Building, 7920 Norfolk Avenue,
Bethesda, 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 20555.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By February 4, 1994, 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 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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 20555, 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 20555, and at the local public document
room for the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: December 2, 1993
Description of amendment requests: The proposed changes would
modify TS 3/4.6.1.2 by removing the schedular requirements for a Type A
(overall integrated containment leakage rate) test to be performed
specifically at 40 1B 10 month intervals and replacing these
requirements with a requirement to perform Type A testing in accordance
with Appendix J to 10 CFR 50.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes would not involve an increase in the
probability or the consequences of an accident previously evaluated.
The proposed change only allows flexibility in the scheduling of the
three required Type A tests in the 10-year service period. The
additional flexibility is needed for plants using 18-month fuel
cycles to allow refueling outages and testing intervals to coincide.
There is no change to the number of tests required, test
methodology, or acceptance criteria.
(2) The proposed changes would not create the possibility of a
new or different type of accident from any accident previously
evaluated. The proposed change to the test schedule only provides
flexibility in meeting the same requirement for three tests in a 10-
year period. The testing type and bases have not changed. Therefore,
operation of the units with this more flexible test schedule will
not result in an accident previously not analyzed in the Updated
Final Safety Analysis Report (UFSAR). The proposed changes do not
impact the design bases of the containment and do not modify the
response of the containment during a design basis accident.
(3) The proposed changes would not involve a reduction in the
margin of safety. The proposed changes to the schedule only provides
flexibility in meeting the same requirement for three tests in a 10-
year period. These proposed changes do not affect or change any
limiting conditions for operation (LCO), or any other surveillance
requirements in the TS, and the basis for the surveillance
requirement remains unchanged. The testing method, acceptance
criteria, and bases are not changed. The TS continue to require
testing that is consistent with the requirements of Appendix J to 10
CFR 50.
The NRC staff has reviewed the licensees' 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
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: 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: Theodore R. Quay
Gulf States Utilities Company, Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 18, 1993
Description of amendment request: The proposed amendment would
permit extending the time to perform leak rate testing of certain
containment isolation valves so that the testing can be performed
during the refueling outage scheduled to start April 16, 1994, rather
than requiring an earlier shutdown solely to perform the testing. The
proposed amendment would revise Surveillance Requirements 4.6.1.3d and
4.6.1.3f to allow a one-time extension of the surveillance intervals
for leak rate testing of containment isolation valves. In addition, the
proposed amendment would revise Surveillance Requirements 4.4.3.2.2a
and 4.4.3.2.2b, replacing the requirement to leak test the reactor
coolant pressure isolation valves every 18 months or prior to returning
a valve to service, with a requirement to leak test the valves in
accordance with the Inservice Testing Program. This would allow the
testing to be performed during the fifth refueling outage.
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. 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:
1. The proposed changes would not significantly increase the
probability or consequences of a previously evaluated accident.
One of the proposed technical specification (TS) changes requests a
one-time only extension of the surveillance intervals for the TS
Surveillance Requirements of TS 4.6.1.3f, leak rate testing of valves
sealed by the main steam positive leakage control system (MS-PLCS) and
the penetration valve leakage control system (PVLCS). The revision
would permit eleven containment isolation valves to be tested a maximum
of 46 days later than required by current technical specifications.
To permit the one-time extension of the surveillance interval for
leak rate tests of containment isolation valves, TS 4.6.1.3d must also
be revised to permit the interval for Type C leak rate tests to exceed
24 months. This change is consistent with an associated exemption
request. The exemption request and this revision would permit 20 valves
to be tested a maximum of 35 days later than required by the current
technical specifications.
The proposed amendment would also revise Surveillance Requirements
4.4.3.2.2a and 4.4.3.2.2b, replacing the requirement to leak test the
reactor coolant pressure isolation valves every 18 months or prior to
returning a valve to service, with a requirement to leak test the
valves in accordance with the Inservice Testing Program. This change
would require that the pressure isolation valves be tested in
accordance with Section XI of the ASME Boiler and Pressure Vessel Code,
resulting in the valves being tested at least every refueling outage,
rather than specifying an 18 month cycle. The revision would permit
five valves to be tested a maximum of 65 days later than allowed under
the current technical specification.
Based on the short duration of the requested extensions, the
extensions will not significantly increase the probability or
consequences of a previously evaluated accident.
2. The proposed changes would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed TS changes permit extension of the surveillance
intervals for leak rate testing of containment isolation valves and
reactor coolant system pressure isolation valves. In that the requested
extension durations are small as compared to the overall interval
allowed by TS, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
3. The proposed changes would not involve a significant reduction
in the margin of safety.
The proposed TS changes permit extension of the surveillance
intervals for leak rate testing of containment isolation valves and
reactor coolant system pressure isolation valves. In that the requested
extension durations are small as compared to the overall interval
allowed by TS, the proposed changes do not involve a significant
reduction in the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: Suzanne C. Black
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: December 8, 1993
Description of amendments request: The frequency for Channel
Calibration would be revised from Q (quarterly) to R (refuel) for
Technical Specification Table 4.3.2.1, Item 4.a.4, High Pressure Core
Injection Steam Line Tunnel Temperature-High.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed change corrects Technical
Specification pages issued for Amendment 166 for Brunswick Unit 1
and Amendment 197 for Brunswick Unit 2, regarding NUMAC Steam Leak
Detection Equipment. Specifically, on page 3/4 3-29 for each unit,
the Channel Calibration frequency of Item 4.a.4, HPCI [High Pressure
Core Injection] Steam Line Tunnel Temperature - High, was
inadvertently left as quarterly (Q) rather than being revised to
refuel (R). The text of CP&L's September 14, 1992 license amendment
request and the NRC's safety evaluation for Amendments 166 and 197,
dated October 14, 1993, addressed the frequency change from
quarterly to refuel for this item. Therefore, the proposed change is
purely administrative in nature and can not involve a significant
increase in the probability of consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, the NRC's safety evaluation for
Amendments 166 and 197, dated October 14, 1993, addressed the
frequency change from quarterly to refuel for Item 4.a.4 of Table
4.3.2-1, HPCI Steam Line Tunnel Temperature - High. Therefore, the
proposed change is purely administrative in nature and can not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in the margin of safety. The proposed change corrects
Technical Specification pages issued for Amendment 166 for Brunswick
Unit 1 and Amendment 197 for Brunswick Unit 2, regarding NUMAC Steam
Leak Detection Equipment. The NRC's safety evaluation for Amendments
166 and 197, dated October 14, 1993, addressed the change of Channel
Calibration frequency of Item 4.a.4, HPCI Steam Line Tunnel
Temperature - High from quarterly (Q) to refuel (R). Therefore, the
proposed change is purely administrative and can not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: S. Singh Bajwa
Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling
Water Reactor (LACBWR), Vernon County, Wisconsin
Date of application for amendment: November 5, 1993 (Reference LAC-
13320)
Brief description of amendment: This proposed change would modify
the Technical Specifications incorporated in Facility Operating License
No. DPR-45 in accordance with the requirements of the revised 10 CFR
Part 20 which becomes mandatory January 1, 1994 (56 FR 23360). In
addition, this proposed change would correct several editorial
oversights from previous amendments.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
the results of its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the information provided by
the licensee and found that the licensee did not provide specific
information as to how it determined that the three standards of
50.92(c) were satisfied. The NRC staff performed its own evaluation of
the proposed change to determine if the three standards of 50.92(c)
were satisfied. The NRC staff's no significant hazards consideration
evaluation is presented below:
1. Will operation of the facility according to this proposed change
involve a significant increase in the probability or consequences of an
accident previously evaluated?
The proposed change is to bring the LACBWR Technical Specifications
into conformance with the revised 10 CFR Part 20 and to correct several
editorial oversights previously evaluated. The proposed change has no
affect on any plant operating parameters. Consequently, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will operation of the facility according to this proposed change
create the possibility of a new or different kind of accident from any
previously evaluated?
The proposed change is to bring the LACBWR Technical Specifications
into conformance with the revised 10 CFR Part 20 and to correct several
editorial oversights previously evaluated. The proposed change is
administrative in nature. Further, the proposed change does not result
in any physical alteration to any plant system, and does not result in
any change in the method by which any safety-related system performs
its function. Consequently, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Will operation of the facility according to this proposed change
involve a significant reduction in a margin of safety?
The margin of safety is the difference between the value of a
critical design, operating, or post accident parameter, and the value
of the parameter which would produce unacceptable results. The proposed
change does not affect any hardware, has no effect on the current
operating methodologies or actions which govern plant performance, and
does not affect any accident analysis parameter. Consequently, the
proposed change does not involve a significant reduction in a margin of
safety.
The NRC staff has determined based on its own no significant
hazards consideration evaluation 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: La Crosse Public Library, 800
Main Street, La Crosse, Wisconsin 54601
Attorney for licensee: Fritz Schubert, Esquire, Dairyland Power
Cooperative, 2615 East Avenue South, La Crosse, Wisconsin 54601
NRC Branch Chief: John H. Austin
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 11, 1993
Description of amendment request: The proposed amendments would
consolidate the Quality Verification Department with the Nuclear
Generation Department and realign the Nuclear Safety Review Board such
that it reports to the Senior Vice-President of the Nuclear Generation
Department.
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 amendments do not involve a significant increase in the
probability or consequences of an accident previously evaluated.]
The proposed revisions to consolidate the Quality Verification
Department with the Nuclear Generation Department and realign the
NSRB [Nuclear Safety Review Board] such that it reports to the
Senior Nuclear Officer, change the reference from Semiannual to
Annual, change the reference from group to division, delete titles
of persons designated to approve modifications, clarify the
responsibilities of the Safety Assurance Manager, and delete the
requirement to perform an annual independent Fire Protection Audit
will not involve a significant increase in the probability or
consequences of an accident previously evaluated because the changes
do not have any impact upon the design or operation of any plant
systems or components.
[2. The amendments do not create the possibility of a new or
different kind of accident from any accident previously evaluated.]
The proposed revisions will not create the possibility of a new
or different kind of accident from any previously evaluated because
the changes are administrative in nature and operation of Catawba,
McGuire, and Oconee Nuclear Stations in accordance with these TS
[technical specifications] will not create any failure modes not
bounded by previously evaluated accidents.
[3. The 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Loren R. Plisco, Acting
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 11, 1993
Description of amendment request: The proposed amendments would
consolidate the Quality Verification Department with the Nuclear
Generation Department and realign the Nuclear Safety Review Board such
that it reports to the Senior Vice-President of the Nuclear Generation
Department.
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 amendments do not involve a significant increase in the
probability or consequences of an accident previously evaluated.]
The proposed revisions to consolidate the Quality Verification
Department with the Nuclear Generation Department and realign the
NSRB [Nuclear Safety Review Board] such that it reports to the
Senior Nuclear Officer, change the reference from Semiannual to
Annual, change the reference from group to division, delete titles
of persons designated to approve modifications, clarify the
responsibilities of the Safety Assurance Manager, and delete the
requirement to perform an annual independent Fire Protection Audit
will not involve a significant increase in the probability or
consequences of an accident previously evaluated because the changes
do not have any impact upon the design or operation of any plant
systems or components.
[2. The amendments do not create the possibility of a new or
different kind of accident from any accident previously evaluated.]
The proposed revisions will not create the possibility of a new
or different kind of accident from any previously evaluated because
the changes are administrative in nature and operation of Catawba,
McGuire, and Oconee Nuclear Stations in accordance with these TS
[technical specifications] will not create any failure modes not
bounded by previously evaluated accidents.
[3. The amendments do not involve a significant reduction in a
margin of safety.]
The proposed revisions will not involve a reduction in a margin
of safety because they are administrative in nature.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Loren R. Plisco, Acting
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: November 11, 1993
Description of amendment request: The proposed amendments would
consolidate the Quality Verification Department with the Nuclear
Generation Department and realign the Nuclear Safety Review Board such
that it reports to the Senior Vice-President of the Nuclear Generation
Department. In addition, the requirement to conduct an annual
independent Fire Protection Audit is deleted.
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 amendments do not involve a significant increase in the
probability or consequences of an accident previously evaluated.]
The proposed revisions to consolidate the Quality Verification
Department with the Nuclear Generation Department and realign the
NSRB [Nuclear Safety Review Board] such that it reports to the
Senior Nuclear Officer, change the reference from Semiannual to
Annual, change the reference from group to division, delete titles
of persons designated to approve modifications, clarify the
responsibilities of the Safety Assurance Manager, and delete the
requirement to perform an annual independent Fire Protection Audit
will not involve a significant increase in the probability or
consequences of an accident previously evaluated because the changes
do not have any impact upon the design or operation of any plant
systems or components.
[2. The amendments do not create the possibility of a new or
different kind of accident from any accident previously evaluated.]
The proposed revisions will not create the possibility of a new
or different kind of accident from any previously evaluated because
the changes are administrative in nature and operation of Catawba,
McGuire, and Oconee Nuclear Stations in accordance with these TS
[technical specifications] will not create any failure modes not
bounded by previously evaluated accidents.
[3. The amendments do not involve a significant reduction in a
margin of safety.]
The proposed revisions will not involve a reduction in a margin
of safety because they are administrative in nature.
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Loren R. Plisco, Acting
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 16, 1993
Description of amendment request: The proposed amendment would
revise the Technical Specifications to change the periodic test
schedule for containment Type A integrated leak rate tests (ILRTs) from
a set of three tests performed at approximately equal intervals during
each 10-year period, as specified in 10 CFR Part 50, Appendix J,
Section III.D, to one Type A test performed at 10-year intervals. The
change is being reviewed in conjunction with a proposed exemption to
Appendix J, as requested by the licensee in a letter dated November 16,
1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The Waterford 3 Type A test history provides substantial
justification for the proposed test schedule. Three type A tests
have been performed over an eight (8) year period with successful
results. The tests indicate that Waterford 3 has a low leakage
containment and that the leakage has never exceeded 24.6% of
La. [La is the maximum allowed leakage rate of air from
containment where containment is pressurized to Pa; for
Waterford 3 Pa is 44 psig. La for Waterford is 0.50
percent by weight of the containment air per 24 hours at Pa.]
There are no structural mechanisms which would adversely affect
the structural capability of the containment and that would be a
factor in extending the Type A test schedule to ten years. A risk
impact assessment was performed, and a determination was made that
there is no risk impact as a result of changing the Type A test
schedule. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
There are no design changes being made that would create a new
type of accident or malfunction. The proposed change will not alter
the plant or the manner in which it is operated. The change proposes
a change to the schedule for performing the periodic Type A test.
The purpose of the test is to provide periodic verification by test
of the leaktight integrity of the primary reactor containment, and
systems and components which penetrate containment. The tests assure
that leakage through containment and systems and components
penetrating containment will not exceed the allowable leakage rate
values associated with conditions resulting from an accident. The
change in schedule for performing the Type A test will not adversely
affect the containment integrity in the event of an accident.
Therefore, the proposed change will not create the possibility of a
new or different type of accident from any accident previously
evaluated.
The proposed change is a change to the schedule for performing
the periodic Type A tests and does not reduce the margin of safety
assumed in accident analysis for release of radioactive materials
from the containment atmosphere into the environment or any margin
of safety preserved by the Technical Specifications. The
methodology, acceptance criteria, and the technical specification
leakage limits for the performance of the Type A tests will not
change, and the Type A tests will be performed in accordance with
10CFR 50, Appendix J, and the Waterford 3 licensing basis.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 16, 1993
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to provide acceptable
conditions for operation when (1) the core operating limits supervisory
system (COLSS) is in service and neither control element assembly
calculator (CEAC) is operable and (2) the COLSS is out of service and
either or both CEACs are operable.
This proposed TS change modifies the departure from nucleate
boiling ratio (DNBR) margin, Limiting Condition for Operation (LCO)
3.2.4b and c, which limits the core power distribution to the initial
value assumed in the accident analyses. Operation within this LCO
either limits or prevents potential fuel cladding failures in the event
of a postulated accident and limits damage to the fuel cladding during
an accident by ensuring that the plant is operating within acceptable
conditions at the onset of a transient. The limiting safety system
settings and this LCO are based on the accident analysis, so that
specified acceptable fuel design limits (SAFDLs) are not exceeded as a
result of anticipated operational occurrences (AOOs) and the limits of
acceptable consequences are not exceeded for other postulated
accidents.
The COLSS and core protection calculators (CPCs) monitor the core
power distribution on line and are capable of verifying that the linear
heat rate (LHR) and DNBR do not exceed their limits. The COLSS performs
this function by continuously monitoring the core power distribution
and calculating core power operating limits corresponding to the
allowable peak LHR and DNBR. The CPCs perform this function by
continuously calculating an actual value of DNBR and LHR for comparison
with the respective trip setpoints. CEACs monitor CEA position. Should
a CEA deviate from its subgroup position, the CEACs will transmit an
appropriate ``penalty'' factor to the CPCs.
The COLSS is normally used to monitor DNBR margin. When at least
one CEAC is operable, TS 3.2.4a provides enough margin to DNB to
accommodate the limiting AOO without failing the fuel. When neither
CEAC is operable, the CPCs lack the CEA position information necessary
to ensure a reactor trip when necessary. In this case TS 3.2.4b
requires the COLSS calculated core power to be reduced to ensure that
the limiting AOO will not result in fuel failure. Currently, TS 3.2.4b
requires that the COLSS calculated power be maintained at 13% below the
COLSS calculated power operating limit to compensate for the potential
error in the CPC DNBR calculation. The proposed revision would increase
this required adjustment to 16%, which is more restrictive than the
present value.
In instances when the COLSS is out of service, but either or both
CEACs are operable, TS 3.2.4c states that the DNBR operating margin
shall be maintained by comparing the DNBR indicated on any operable CPC
channel with the allowable value from TS Figure 3.2-2. Whenever the
COLSS is out of service, the CPCs are used to perform the same
monitoring function. However, the extra conservatisms built into the
CPCs for transient protection are not all required when the CPCs are
being used for monitoring. In order not to affect the CPC transient
protection, these conservatisms are not taken from the CPC, but are
credited in the COLSS out-of-service limits in Figure 3.2-2. A
reevaluation of the limiting AOOs has verified that, by maintaining the
margin in the proposed Figure 3.2-2, sufficient margin exists to ensure
that the limiting Cycle 7 AOO will not result in fuel failure.
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:
For the case when neither CEAC is operable but COLSS is in
service, the CPCs assume a preset CEA configuration and can not
obtain the required CEA position information to ensure the SAFDL on
DNBR will not be violated during an AOO. Thus, as a result of
limiting AOOs for Cycle 7, Specification 3.2.4b requires that core
power be reduced to a value 16% less than the current COLSS
calculated power operating limit. This ensures the limiting AOO will
not result in a violation of SAFDLs. The proposed revision to Figure
3.2-2 accounts for the situation when COLSS is out of service but at
least one CEAC is operable. In this case, the Cycle 7 safety
analysis has shown that by maintaining the CPC calculated DNBR above
the value shown in the figure, the limiting AOO will not result in a
violation of the SAFDLs. Therefore, the proposed change will not
significantly increase the probability or consequences of any
accident previously evaluated.
The proposed changes are primarily a result of changes in Cycle
7 core parameters. These changes do not involve any change to any
equipment or manner in which the plant will be operated. These
changes further restrict the plant operation when either COLSS or
both CEACs are out of service. Therefore, the proposed change will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The intent of this Specification is to ensure that there is
always sufficient margin to DNB such that the CPCs can mitigate the
consequences of the most limiting AOO prior to a violation of the
SAFDLs. Generally, this margin is continuously monitored by COLSS;
however, if COLSS is out of service, but at least one CEAC is
operable, the limitation on CPC calculated DNBR (as a function of
ASI [axial shape index]) shown in Figure 3.2-2 represents a
conservative envelope of operating conditions consistent with the
Cycle 7 safety analysis assumptions. This band of operating
conditions has been analytically demonstrated to maintain an
acceptable minimum DNBR through all AOOs. On the other hand, for the
case when COLSS is in service, but neither CEAC is operable, the
proposed change will ensure that the limiting AOO will not result in
a violation of SAFDLs. Therefore, the proposed changes 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: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: December 2, 1993
Description of amendment request: The purpose of the request is to
change the plant Technical Specifications (TS) to remove the limiting
conditions for operation and surveillance requirements for the chlorine
detection system. TMI-1 removed the gases Chlorination System for the
Circulating Water and River Water Systems. This modification eliminated
the need for a Chlorine Dectection System (CDS).
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. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The TS requirements assured the operability of
the CDS in the event of an on-site chlorine release from a one ton
cylinder. These TS requirements reduced the probability and the
consequences of a radiological accident which may result from an
incapacitation of control room operatorsafter entry of chlorine into
the control room. With the removal and the restriction on delivery
of one ton chlorine cylinders, this postulated event is no longer
credible, and there is a decrease in the probability of a
radiological accident.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The TS
requirements associated with the CDS were for the on-site release of
chlorine from a one ton cylinder. These cylinders are removed and
prohibited from the TMI-1 site. These actions preclude a significant
on-site release of chlorine which could affect the control room
operators.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The purpose of the TS requirements was to maintain
operability of the CDS in the event of on-site release from a one
ton chlorine cylinder. Since chlorine cylinders greater than 150
pounds are prohibited on-site, the TS requirements for chlorine
detection are no longer required, and their removal will not reduce
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 Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: November 23, 1993
Description of amendment request: The licensee proposes to modify
the South Texas Project, Units 1 and 2, Technical Specification 3/
4.8.1.1, ``A.C. Sources,'' to modify the action statements and
surveillance requirements for testing of the standby diesel generator.
This amendment would incorporate the recommendations of NRC Generic
Letter (GL) 93-05, ``Line-Item Technical Specifications Improvements To
Reduce Surveillance Requirements For Testing During Power Operation,''
dated September 27, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change seeks to eliminate the unnecessary testing
of an operable Standby Diesel Generator (SDG). Technical
Specification (TS) 3.8.1.1 Actions a. and e. require all operable
SDGs be started as a demonstration of operability whenever one or
more of the offsite AC [alternating current] power sources is
declared inoperable. The inoperability of an offsite AC power source
has no effect on the reliability of a SDG. Deleting this requirement
does not affect the design or performance characteristics of the
SDGs. Therefore, the SDGs maintain their ability to perform their
design function.
TS 3.8.1.1 Actions b. and c. require all remaining operable SDGs
be started as a demonstration of operability whenever one of the SDG
is declared inoperable except for preplanned preventive maintenance
or testing. The proposed amendment would expand the testing
exclusion to include an inoperable support system and an
independently testable component in addition to preplanned
preventive maintenance and testing. The proposed amendment would
also eliminate the testing requirement of the remaining operable
SDGs, when a SDG is declared inoperable, unless there is cause to
believe a potential common mode failure exists for the remaining
SDGs. The normal TS surveillance testing schedule assures that
operable SDG(s) are capable of performing their intended safety
functions. A failure of one SDG does not reduce the reliability of
another, otherwise operable SDG. Deleting this requirement does not
affect the design or performance characteristics of the SDGs, once a
common mode failure has been dismissed. Therefore, the SDGs maintain
their ability to perform their design function.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The elimination of these unnecessary tests does not affect the
design bases of the SDGs, or any of the accident evaluations
involving the SDGs. The SDGs are designed to provide electrical
power to the equipment important for safety during all modes and
plant conditions following a loss of offsite power. The test
schedule established in accordance with GL 84-15 [``Proposed Staff
Actions To Improve and Maintain Diesel Generator Reliability'']
assures that operable SDGs are capable of performing their intended
safety function. Therefore, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
Since the proposed change does not affect the design bases,
accident analysis, reliability or capability of the SDGs to perform
their intended safety function, this change does not involve any
reduction in a margin to 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
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, NW, Washington, DC 20036
NRC Project Director: Suzanne C. Black
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of amendment request: December 15, 1993. This submittal
supersedes a previous submittal dated March 10, 1993.
Description of amendment request: The proposed amendment would
implement interim tube plugging criteria for the tube support plate
elevation outer diameter stress corrosion cracking for cycle 14.
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. Operation of Donald C. Cook Nuclear Plant Unit 1 in
accordance with the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Testing of model boiler specimens for free span tubing (no TSP
restraint) at room temperature conditions shows burst pressures in
excess of 5000 psi for indications of ODSCC with voltage
measurements as high as 19 volts. Burst testing performed on pulled
tubes from Cook Nuclear Plant Unit 1 with up to a 2.02 volt
indication shows measured burst in excess of 10,000 psi at room
temperature. Correcting for the effects of temperature on material
properties and minimum strength levels (as the burst testing was
done at room temperature), tube burst capability significantly
exceeds the RG 1.121 criterion requiring the maintenance of a margin
of 3 times normal operating pressure differential on tube burst. The
3 times normal operating pressure differential for the Cook Nuclear
Plant Unit 1 steam generators corresponds to 4275 psi. Based on the
existing data base, this criterion is satisfied with 7/8'' diameter
tubing with bobbin coil indications with signal amplitudes less than
4.9 volts, regardless of the indicated depth measurement. A 1.0 volt
plugging criteria compares favorably with the structural limit
considering the previously calculated growth rates for ODSCC within
the Cook Nuclear Plant Unit 1 steam generators. Considering a
voltage increase of 0.4 volts, and adding a 20% NDE uncertainty of
0.20 volts (90% Cumulative Probability) to the IPC of 1.0 volts
results in an EOC voltage of approximately 1.6 volts for Cycle 14
operation. A 3.3 volt safety margins implied (4.9 structural limit -
1.6 volt EOC - 3.3 volt margin). This EOC voltage compares favorably
with the Structural Limit of 4.9 volts.
For the voltage/burst correlation, the EOC structural limit is
supported by a voltage of 4.9 volts. A 3.1 volt BOC repair limit
confirms the structural limit when 40% growth and 20% uncertainty
are applied to the repair limit. This repair limit will be applied
for Cycle 14 IPC implementation to repair bobbin indication greater
than 3.1 volts independent of RPC confirmation of the indication.
The conservatism of this repair limit is shown by the EOC 12
(Summer 1992) eddy current data. The overall average voltage growth
was determined to be only 2.2%, with a 12% average voltage growth
for indications less than 0.75 volt BOC and a 1% average voltage
growth for indication >0.75 volt at the BOC. In addition, the Cycle
12 maximum observed voltage increase was found to be 0.49 volts, and
occurred in a tube initially <1.0 BOC. In accordance with the
technical specification requirements, the applicability of Cycle 13
growth rates for Cycle 14 operation will be confirmed prior to
return to power of Cook Nuclear Plant Unit 1. Similar large
structural margins are anticipated.
As stated previously, TSP proximity to the tubes will prevent
tube burst during all plant conditions. Test data indicates that
tube burst cannot occur within the TSP, even for tubes which have
100% through-wall EDM notches, 0.75 inch long, provided that the TSP
is adjacent to the notched area. Therefore, a more realistic
assessment of tube operability should be performed against the RG
1.121 loading requirements during accidents SLB conditions, since
the TSP has the potential to deflect during blowdown following a
main SLB, thereby uncovering the intersection. At the ASME Code
recommended faulted condition loading of 3657 psi (2560 psi/0.7)
structural integrity is provided for bobbin voltage indications of a
minimum of 9.6 volts. The repair limit based on the structural
limited conservative SLB conditions would be 6.0 volts (compared to
a 3.1 volt repair limit for a structural limit based on the
3[delta]P burst capability voltage).
Only three indications of ODSCC have been reported to have
operating leakage, and all three have been in European plants. No
field leakage has been reported at other plants from tubes with
indications of a voltage level of under 7.7 volts (from 3/4''
tubing). Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main
SLB outside of containment but upstream of the MSIV represents the
most limiting radiological condition relative to the IPC. In support
of implementation of the IPC, it will be determined whether the
distribution of cracking indications at the TSP intersections at the
EOC 14 are projected to be such that primary to secondary leakage
would result in site boundary doses within a small fraction of the
10 CFR 100 guidelines. The SLB leakage rate calculation methodology
prescribed in Section 3.3 of draft NUREG-1477 will be used to
calculate EOC 14 leakage. Due to the relatively low voltage growth
rates at Cook Nuclear Plant Unit 1 and the relatively small number
of indications affected by the IPC, SLB leakage prediction per draft
NUREG-1477 is expected to be less than the acceptance limit of 1.0
gpm in the faulted loop and far below the conservatively calculated
SRP based allowable value of 120 gpm in the faulted loop. The NRC
leakage rate calculation methodology applies a 98% confidence limit
on leakage that is independent of voltage. This method for
calculating SLB leakage is conservative as it assumes no
correlations exists between SLB leakage and bobbin probe voltage.
Tube pull results from Cook Nuclear Plant Unit 1 indicate that tube
wall degradation of greater than 40% through-wall was detectable
either by the bobbin or RPC probe. The tube with maximum through-
wall penetration of 56% (42% average) had a voltage of 2.02 volts.
This indication also was the largest recorded bobbin voltage from
the EOC 12 eddy current data. All burst tested tube intersections
had degradation depths of 40% to 56 % (actual) deep and all were
detected by both probes, with all bobbin voltage grater than or
equal to 1.0. Since the criteria requires the plugging of >1.0 volt
bobbin indications with confirmed RPC calls, using the Cook Nuclear
Plant Unit 1 pulled tube destructive examination results, it is
reasonable that no indications of degradation greater than 40% to
56% deep with an ability to influence tube burst capability were
left in service. Since the majority of the EOC 14 indications at
Cook Nuclear Unit 1 are expected to be below this level, the
inclusion of all IPC intersections into the leakage calculation is
exceptionally conservative.
Therefore, as re-implementation of the 1.0 volt IPC during Cycle
14 does not adversely affect steam generator tube integrity and
results in acceptable dose consequences, the proposed amendment does
not result in any increase in the probability or consequences of an
accident previously evaluated within the Cook Nuclear Plant Unit 1
FSAR.
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 proposed steam generator tube IPC does not
introduce any significant changes to the plant design basis. Use of
the criteria does not provide a mechanism which could result in a
tube rupture outside of the region of the TSP elevations; no ODSCC
is occurring outside the thickness of the TSPs. Neither a single or
multiple tube rupture event would be expected in a steam generator
in which the plugging criteria has been applied (during all plant
conditions).
Specifically, Cook Nuclear Plant will continue to implement a
maximum leakage rate limit of 150 gpd (0.1 gpm) per steam generator
to help preclude the potential for excessive leakage during all
plant conditions. The Cycle 14 Technical Specification limits on
primary to secondary leakage at operating conditions is a maximum of
0.4 gpm (600 gpd) for all steam generators, or, a maximum of 150 gpd
for any one steam generator. The RG 1.121 criterion for establishing
operational leakage rate limits that require plant shutdown are
based upon leaks-before-break consideration to detect a free span
crack before potential tube rupture. The 150 gpd limit should
provide for leakage detection and plant shutdown in the event of the
occurrence of an unexpected single crack resulting in leakage that
is associated with the longest permissible crack length. RG 1.121
acceptance criteria for establishing operating leakage limits are
based on leak-before-break considerations such that plant shutdown
is initiated if the leakage associated with the longest permissible
crack is exceeded. The longest permissible crack is the length that
provides a safety factor of 3 against bursting at normal operating
pressure differential. A voltage amplitude of 4.9 volts for typical
ODSCC corresponds to meeting this tube burst requirement at a lower
95% prediction limit on the burst correlation coupled with 95/95 LTL
material properties. Alternate crack morphologies can correspond to
4.9 volts so that a unique crack length is not defied by the burst
pressure versus voltage correlation. Consequently, typical burst
pressure versus through-wall crack length correlations are used
below to define the ``longest permissible crack'' for evaluating
operating leakage limits.
At current plant conditions, the single through-wall crack
lengths that result in tube burst at 3 times normal operating
pressure differential and SLB conditions are 0.44 inch and 0.84
inch, respectively. A leak rate of 150 gpd will provide for
detection of 0.42 inch long cracks at nominal leak rates and 0.61
inch long cracks at the lower 95% confidence level leak rates. Since
tube burst is precluded during normal operation due to the proximity
of the TSP to the tube and the potential for the crevice to become
uncovered during SLB conditions, the leakage from the maximum
permissible crack must preclude tube burst at SLB conditions. Thus,
the 150 gpd limit provides for plant shutdown prior to reaching
critical crack lengths for SLB conditions.
3. The proposed license amendment does not involve a significant
reduction in margin of safety.
The use of the voltage based bobbin probe interim TSP elevation
plugging criteria at Cook Nuclear Plant Unit 1 is demonstrated to
maintain steam generator tube integrity commensurate with the
criteria of Regulatory Guide 1.121. RG 1.121 describes a method
acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by
reducing the probability or the consequences of steam generator tube
rupture. This is accomplished by determining the limiting conditions
of degradation of steam generator tubing, as established by
inservice inspection, for which tubes with unacceptable cracking
should be removed from service. Upon implementation of the criteria,
even under the worst case conditions, the occurrence of ODSCC at the
TSP elevations is not expected to lead to a steam generator tube
rupture event during normal or faulted plant conditions. The EOC 14
distribution of crack indications at the TSP elevations will be
confirmed to result in acceptable primary to secondary leakage
during all plant conditions and that radiological consequences are
not adversely impacted.
In addressing the combined effects of LOCA + SSE on the steam
generator component (as required by GDC 2), it has been determined
that tube collapse may occur in the steam generators at some plants.
This is the case as the TSPs may become deformed as a result of
lateral loads at the wedge supports at the periphery of the plant
due to the combined effects of the LOCA rarefaction wave and SSE
loadings. Then, the resulting pressure differential on the deformed
tubes may cause some of the tubes to collapse.
There are two issues associated with steam generator tube
collapse. First, the collapse of steam generator tubing reduces the
RCS flow area through the tubes. The reduction in flow are increases
the resistance to flow of steam from the core during a LOCA which,
in turn, may potentially increase peak clad temperature (PCT).
Second, there is a potential that partial through-wall cracks in
tubes could progress to through-wall cracks during tube deformation
or collapse.
Consequently, since the leak-before-break methodology is
applicable to the Cook Nuclear Plant Unit 1 reactor coolant loop
piping, the probability of breaks in the primary loop piping is
sufficiently low that they need not be considered in the structural
design of the plant. The limiting LOCA event becomes either the
accumulator line brake or the pressurizer surge line break. LOCA
loads for the primary pipe breaks were used to bound the Cook
Nuclear Plant Unit 1 smaller breaks. The results of the analysis
using the larger break inputs show that the LOCA loads were found to
be of insufficient magnitude to result in steam generator tube
collapse or significant deformation.
Addressing RG 1.83 consideration, implementation of the bobbin
probe voltage based interim tube plugging criteria of 1.0 volt is
supplemented by the following: enhanced eddy current inspection
guidelines to proved consistency in voltage normalization, a 100%
eddy current inspection sample size at the TSP elevations, and RPC
inspection requirements as outlined in the technical specifications
and Appendix A ``NDE Data Acquisition and Analysis Guidelines''
(Attachment 6).
As noted previously, implementation of the TSP elevation
plugging criteria will decrease the number of tubes which must be
repaired. The installation of steam generator tube plugs reduce the
RCS flow margin. Thus, implementation of the alternate plugging
criteria will maintain the margin of flow that would otherwise be
reduced in the event of increased tube plugging.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the Final Safety
Analysis Report or any of the plant Technical Specifications.
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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Leaks Preston Palenske
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: A Randolph Blough, Acting
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 amendment requests: November 15, 1993
Description of amendment requests: The proposed amendments delete
certain Limiting Conditions for Operation, Actions, and Surveillance
Requirements for Reactor Coolant System Pressure Isolation Valves in
the Technical Specifications. The Technical Specifications for these
Reactor Coolant System Pressure Isolation Valves were added by Order
dated April 20, 1981. This Order was prompted by concerns for an
interfacing system loss-of-coolant accident as identified in the
Reactor Safety Study (WASH-1400). The proposed Technical Specification
change, by inference, also requests rescission of the April 20, 1981
Order.
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:
Per 10 CFR 50.92, a proposed amendment to an operating license
will not involve a significant hazards consideration if the proposed
amendment satisfies the following three criteria:
1. Does not involve a significant increase in the probability or
consequences of an accident previously analyzed,
2. Does not create the possibility of a new or different kind of
accident from an accident previously analyzed or evaluated, or
3. Does not involve a significant reduction in a margin of
safety.
Criterion 1
The ISLOCA is not one of the accidents previously analyzed in
Chapter 14, Safety Analysis, of the Cook Nuclear Plant Updated Final
Safety Analysis Report. Chapter 14 analyzes the large break LOCA in
Section 14.3.1, and ``loss of reactor coolant from small ruptured
pipes or from cracks in large pipes which actuates the ECCS'', or
small break LOCA in Section 14.3.2. Therefore, deleting from the
Technical Specifications the Reactor Coolant System pressure
isolation valves in Table 3.4-0, will not increase the probability
or the consequences of the large break or the small break LOCAs
previously analyzed for the Cook Nuclear Plant.
Criterion 2
The Reactor Coolant System pressure isolation valves in Table
3.4-0 of the Technical Specifications were added because WASH-1400
identified the ISLOCA as a significant contributor to core damage
frequency. Deletion of the subject valves from the Technical
Specifications and reliance on the testing requirements mandated by
the In-Service Testing Program of ASME XI does not create the
possibility of a new or different kind of accident from the large
break or the small break LOCAs previously analyzed for the Cook
Nuclear Plant.
Criterion 3
Deleting the Reactor Coolant System pressure isolation valves
from the testing requirements in Table 3.4-0 of the Technical
Specifications will result in these valves only being tested on a
refueling outage frequency as part of the ASME B&PV Code Section XI
IST Program. This somewhat reduced testing frequency will result in
a slight increase in the ISLOCA contribution to core damage
frequency of 5.4%, from lower 5.00E-08/reactor year to mid 5.00E-08/
reactor year. This insignificant increase will not affect the
overall core damage frequency of 6.26E-05/reactor year. Therefore,
it is concluded that the proposed deletion of the Reactor Coolant
System pressure isolation valves in Table 3.4-0 of the Technical
Specifications, as well as the proposed deletion of the portions of
the Technical Specifications that are affected by Table 3.4-0, will
not result in a significant reduction in the margin of safety that
exists at Cook Nuclear Plant to prevent an ISLOCA or to mitigate the
consequences of an ISLOCA.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: A. Randolph Blough, Acting
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay
Power Plant, Unit 3, Humboldt County, California
Date of application for amendment: October 8, 1993 (Reference LAR
93-02)
Brief description of amendment: This Licensee Amendment Request
(LAR) proposes to revise the Humboldt Bay Power Plant (HBPP), Unit 3,
Technical Specifications (TS) by deleting Figure II-2 in Section II,
``Site,'' by deleting the Restricted Area boundary line in Figure V-3,
Section V, ``Monitoring Systems,'' by incorporating a title change into
Section VII, ``Administrative Controls,'' and by revising Figure VII-2,
``Plant Staff Organization.'' The proposed changes are in response to
the revised 10 CFR Part 20 which becomes mandatory on January 1, 1994
(56 FR 23360). The specific TS changes proposed are as follows:
(1) Page v, Figures - delete reference to Figure II-2.
(2) Page II-1, Section II.B, Plant Areas - change ``is shown in
Figure II-2'' to ``shall be defined in plant procedures.''
(3) Page II-3, Section II - delete Figure II-2.
(4) Page V-14, Section V - delete the Restricted Area boundary line
from Figure V-3, ``HBPP Groundwater Monitoring Systems Wells,'' to be
consistent with item 3 above.
(5) Page VII-5, Section VII.C.2.e, Supervisor of Maintenance -
change the title from ``Supervisor of Maintenance'' to ``Maintenance
Planner.''
(6) Page VII-10, Section VII.D.1.b., Membership, List of minimum
membership - replace ``Supervisor of Maintenance'' with ``Maintenance
Planner.''
(7) Page VII-31, Section VII, Figure VII-2, Plant Staff
Organization - replace ``Maintenance Supervisor'' with ``Maintenance
Planner.''
(8) Page VII-31, Section VII, Figure VII-2, Plant Staff
Organization - both the Mechanical Foreman and the Instrument/
Electrical Foreman report directly to the Plant Manager, not to the
``Maintenance Planner,'' as previously shown.
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:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
A change to the defined restricted area has no affect on any
plant operating parameters. Consequently, a change to the defined
restricted area will not affect the probability or consequences of
an accident occurring.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed revisions to the HBPP TS are administrative in
nature. Further, the proposed changes would not result in any
physical alteration to any plant system, and there would not be a
change in the method by which any safety-related system performs its
function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The proposed revisions to the HBPP TS do not affect the margin
of safety of any accident analysis since they do not affect the
parameters for any accident analysis, and have no effect on the
current operating methodologies or actions which govern plant
performance.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas &
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Branch Chief: John H. Austin
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: October 29, 1993
Description of amendment request: The amendment would revise the
Limerick Generating Station, Units 1 and 2, Technical Specifications to
eliminate the Main Steam Line Radiation Monitoring System high
radiation trip function for initiating 1) an automatic reactor scram
and automatic closure of the Main Steam Line Isolation Valves, and 2)
automatic closure of the Main Steam Line drain valves, and Main Steam
and Reactor Water Sample line valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specification (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes involve eliminating the Main Steam Line
Radiation Monitoring (MSLRM) system high radiation trip function for
initiating an automatic reactor scram and automatic closure of the
Main Steam Line Isolation Valves (MSIVs), Main Steam line drain
valves, and Main Steam and Reactor Water sample line valves. The
proposed TS changes support installation of a plant modification to
defeat portions of MSLRM system high radiation trip function logic
circuitry in the Reactor Protection System (RPS) and Primary
Containment and Reactor Vessel Isolation Control System (PCRVICS).
Installation of this modification will not adversely impact the
operation of the RPS or PCRVICS with respect to performing its other
intended safety functions. The proposed TS changes will not affect
the operation of other plant systems or equipment important to
safety. The MSLRM system high radiation trip function for the
Mechanical Vacuum Pump (MVP) will be retained. The safety assessment
and justification for eliminating the MSLRM system high radiation
trip function for initiating an automatic reactor scram and
automatic closure of the MSIVs [are] based on General Electric's
(GE's) Topical Report NEDO-31400A, ``Safety Evaluation for
Eliminating the Boiling Water Reactor Main Steam Line Isolation
Valve Closure Function and Scram Function of the Main Steam Line
Radiation Monitor,'' and the applicability of this report to
Limerick Generating Station (LGS), Units 1 and 2. By letter dated
May 15, 1991, the NRC approved this topical report and indicated
that it was acceptable for licensees to reference this report as the
basis for requesting a TS change to eliminate the MSLRM system high
radiation trip functions as documented in the report and associated
NRC Safety Evaluation Report (SER).
The safety assessment provided in NEDO-31400A can also be
applied to eliminate the MSLRM system high radiation trip function
for initiating the automatic closure of the Main Steam line drain
valves although this aspect was not explicitly evaluated in NEDO-
31400A. The flow from these valves ultimately discharges to the main
condenser as do the MSIVs and therefore, any radioactive material
passing through these valves would be processed in the same fashion
as that passing through the MSIVs. The effects of eliminating the
MSLRM system high radiation trip function for initiating the closure
of the Main Steam and Reactor Water sample line valves is [are]
negligible. The sample lines are routed to a sample sink where inlet
valves installed on the sample lines are normally closed.
Additionally, downstream of the inlet valves are needle valves
designed to control and limit sample line flow. The sample sink is
enclosed, and air vented from its exhaust hood is passed through
filters prior to release to the environment. There is the potential
that a minimal amount of radioactive material could be released to
the environment if the sample sink inlet and needle valves failed to
properly function. This potential release has been evaluated and
determined to a small fraction of the dose limit requirements
specified in 10 CFR 100.
The MSLRM system high radiation trip was intended to function in
response to a Control Rod Drop Accident (CRDA), a Design Basis
Accident previously evaluated. Although the CRDA assumes MSIV
closure, no credit was taken for this in the CRDA analysis since it
postulates that the radioactive material calculated to be released
from the fuel is transported to the main condenser prior to the
MSIVs completely closing. Furthermore, the probability of a fuel
failure is independent of the operation of the MSLRM system.
The Steam Jet Air Ejectors (SJAEs) will continue to operate to
remove non-condensable gases from the main condenser for processing
by the Offgas Treatment system. The Offgas Treatment system will
continue to function as designed to reduce offgas radioactivity
levels prior to release to the environment. Eliminating the MSLRM
system high radiation isolation functions will improve operational
flexibility in that the main condenser will be available to aid in
decay heat removal. Elimination of the MSLRM system high radiation
trip functions in conjunction with proper operation of the Offgas
Treatment system will ensure that any radioactive material released
to the environment is a small fraction of 10 CFR 100 limits.
Therefore, the proposed TS changes associated with eliminating
the MSLRM system high radiation trip function for initiating an
automatic reactor scram and automatic closure of the MSIVs, Main
Steam line drain valves, and Main Steam and Reactor Water sample
line valves do not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes involve eliminating the MSLRM system
high radiation trip function for initiating an automatic reactor
scram and automatic closure of the MSIVs, Main Steam Line drain
valves, and Main Steam and Reactor Water sample line valves. The
proposed TS changes will not affect the operation of other plant
systems or equipment important to safety. The associated plant
modification simply defeats the MSLRM system high radiation trip
function logic circuitry in the RPS and PCRVICS. The RPS and PCRVICS
will continue to respond in performing its other design intended
safety functions. The MSLRM system high radiation trip function for
the MVP will be retained. The proposed TS changes do not involve any
plant hardware changes that could introduce any new failure modes or
effects. The MSLRM system radiation monitors will remain active to
initiate Main Control Room (MCR) annunciation alarms. Plant
procedures will be in place to implement the appropriate mitigative
measures in response to a MSLRM system high radiation alarm signal.
The SJAEs will continue to operate to remove non-condensable
gases from the main condenser for processing by the Offgas Treatment
system. The Offgas Treatment system will continue to function as
designed to reduce offgas radioactivity levels prior to release to
the environment.
Since the Design Basis Accident analysis (i.e., CRDA) does not
credit the MSLRM system high radiation trip function for reducing
the radiological consequences of the postulated accident, the
proposed TS changes have effectively been evaluated and are included
in the existing analysis. That is, the CRDA analysis already assumes
that the radioactive material released from the failed fuel is
immediately transported to the main condenser prior to the MSIVs
completely closing.
The safety assessment and assumptions documents in GE Topical
Report NEDO-31400A provide the basis for eliminating the MSLRM
system high radiation trip function for initiating an automatic
reactor scram and automatic closure of the MSIVs. The safety
assessment provided in NEDO-31400A can also be applied to eliminate
the MSLRM system high radiation trip function for initiating the
closure of the Main Steam Line drain valves, since any radioactive
material passing through these valves would be processed in the same
fashion as that passing through the MSIVs. Eliminating the MSLRM
system high radiation trip function for initiating the closure of
the Main Steam and Reactor Water sample line valves will have a
negligible impact. The sample lines are routed to a sample sink
where inlet valves installed on the sample lines are normally
closed. Downstream of the inlet valves are needle valves designed to
control and limit sample line flow. The sample sink is located in
the Reactor Enclosure and is enclosed, and air vented from its
exhaust hood is passed through filters prior to release to the
environment. The Reactor Enclosure ventilation duct radiation
monitor samples air from the sample sink hood exhaust, and will
isolate the Reactor Enclosure ventilation system if the radiation
levels exceed the monitor's setpoint. There is the potential that a
minimal amount of radioactive material could be released to the
environment through this flowpath if the sample sink inlet and
needle valves failed to properly function. This potential release
has been evaluated and determined to a small fraction of the dose
limit requirements specified in 10CFR100.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes to eliminate the MSLRM system high
radiation trip function for initiating an automatic reactor scram
and automatic closure of the MSIVs, Main Steam line drains valves,
and Main Steam and Reactor Water sample line valves do not change
the conclusion reached in the LGS Updated Final Safety Analysis
Report (UFSAR) that the calculated radiological consequences of the
bounding Design Basis Accident (i.e., CRDA) will not exceed the dose
limit requirements established by 10 CFR 100. The proposed TS
changes will improve the overall reliability of the plant when
compared to the existing system lineup configuration, since it will
reduce the potential of an unnecessary plant transient occurring as
a result of an inadvertent MSIV closure.
A reliability assessment analysis was performed to evaluate the
effects of eliminating the MSLRM system high radiation reactor scram
function on reactivity control failure frequency and core damage
frequency in GE Topical Report NEDO-31400A. This analysis indicated
that there is a negligible increase in reactivity control frequency
with the elimination of the MSLRM trip function. However, this
increase is compensated for by the reduction in transient initiating
events (i.e., inadvertent reactor scrams). This reduction in
transient initiating events represents a reduction in core damage
frequency and thus, results in a net improvement in safety.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Larry E. Nicholson, Acting
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: November 30, 1993
Description of amendment request: The amendment would extend the
surveillance interval of the primary containment drywell-to-suppression
chamber bypass leak test from the current 18-month interval as required
by Technical Specification (TS) Surveillance Requirement 4.6.2.1.d to a
40 +/- 10-month interval. This change would allow the drywell-to-
suppression chamber bypass test to coincide with the 10 CFR 50,
Appendix J, Type A test (i.e., Containment Integrated Leakage Rate Test
(CILRT)) interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The failure effects that are potentially created by the proposed
Technical Specifications (TS) changes have been considered. The
accident which is potentially negatively impacted by the proposed TS
changes are any Loss of Coolant Accident (LOCA) inside primary
containment with or without offsite power available.
The proposed TS changes increase the surveillance interval of
the drywell-to-suppression chamber bypass leak test required by TS
Section 4.6.2.1.d, and will require that an additional test be
performed on the downcomer vacuum breakers assemblies. The primary
containment structure and associated equipment are not considered to
be accident initiators, they act to mitigate the consequences of an
accident. There are no physical or operational changes being made as
a result of these proposed changes. Therefore, the probability of
occurrence of an accident previously evaluated is not increased.
There is a potential increased risk that an increase in the
bypass leakage may go undetected for the duration of the proposed
extension of the interval between the performance of the drywell-to-
suppression chamber bypass leak test. However, as discussed below,
the increased risk is considered to be negligible due to the design
of the diaphragm structure and past test data. Therefore, we have
concluded that the probability of bypass leakage exceeding the
allowed value is not increased as a result of the proposed TS
changes.
The proposed TS changes will extend the surveillance interval
for the drywell-to-suppression chamber bypass leak test from 18
months to 40 +/- 10 months. These proposed changes would allow this
test to be performed at the same interval as the 10CFR50, Appendix
J, Type A test (i.e., Containment Integrated Leakage Rate Test
(CILRT)). In addition, the proposed changes will add an additional
surveillance requirement to be performed on the vacuum breaker
assemblies during refueling outages when the drywell-to-suppression
chamber bypass leak test is not required to be performed. The
proposed TS changes do not increase the consequences of an accident
previously evaluated. This is based on the evaluation summarized
below that demonstrates that the overall impact, if any, on the
plant containment integrity is negligible. Furthermore, the
performance history for the previous LGS bypass leak tests does not
indicate any time based failures. The proposed TS changes also
include a change to the frequency of testing, if two consecutive
tests fail, from once every nine (9) months to once every 24 months
in order to coincide with the 24 month refueling cycle. This change
has no impact on the consequences of an accident based on
maintaining the original requirement to increase the frequency of
testing if two consecutive bypass leak tests fail, and maintaining a
TS requirement for the NRC to review the schedule for subsequent
tests.
During a LOCA inside containment, potential leak paths between
the drywell and suppression chamber airspace could result in
excessive containment pressures, since the steam flow into the
airspace would bypass the heat sink capabilities of the suppression
pool. The containment pressure response to the postulated bypass
leakage can be mitigated by manually actuating the suppression
chamber sprays. Accordingly, since the sprays are manually actuated,
an analysis was performed to show that the operator has sufficient
time to initiate the sprays prior to exceeding the containment
design pressure. This analysis is described in section 6.2.1.1.5 of
the LGS Updated Final Safety Analysis Report (UFSAR). The analysis
is based on a small break LOCA inside containment with a
differential pressure between the drywell-to-suppression chamber
equal to the static pressure due to downcomer submergence. The
analysis concludes that the containment design pressure of 55 psig
will be reached in over 30 minutes from the onset of a small break
LOCA assuming a drywell-to-suppression chamber bypass flow area
(i.e., A/square root of k) equal to 7.20 in2 without operator
intervention.
TS Limiting Condition for Operation 3.6.2.1.b conservatively
specifies a maximum allowable bypass area of 10 % of the design
value of 7.20 in2. This TS limit provides an additional safety
factor of 10 above the conservatism taken in the steam bypass
analysis (i.e., 0.720 in2). The drywell-to-suppression chamber
bypass leak test required by TS Surveillance Requirement 4.6.2.1.d
verifies that the actual bypass flow area is less than or equal to
the TS limit of 0.720 in2. The bypass leakage test ensures that
degradation in the measured bypass area is identified and corrected
to ensure containment integrity during LOCA events.
The potential bypass leakage paths can be divided into two
categories as described below.
1) Leakage pathways other than those associated with the
drywell-to-suppression chamber vacuum breaker assemblies such as
diaphragm floor penetrations (i.e., downcomer and Safety/Relief
Valve (SRV) discharge line penetrations), cracks in the diaphragm
floor and/or liner plate, and cracks in the downcomers and SRV
discharge lines that pass through the suppression chamber airspace.
2) The four sets of drywell-to-suppression chamber vacuum
breaker assemblies.
All other potential bypass leakage pathways have at least two
isolation valves in the potential leakage path. These valves are
high quality leak-tight containment isolation valves that are
normally closed and receive an isolation signal to close. All Air
Operated Valves (AOVs) in these paths fail closed.
Several plant design features and the bypass leak test data
measured to date confirm that the leakage from other than the vacuum
breaker assemblies is negligible and indicates that this leakage
will continue to be negligible for the proposed increased duration
between tests. All pressure boundary penetrations between the
drywell and the suppression chamber are welded except the vacuum
breaker valves and the blind flanges closing 10 spare nozzles in the
downcomers. All pressure boundary penetrations between the drywell-
to-suppression chamber have been fabricated, erected, and inspected
in accordance with the American Society of Mechanical Engineers
(ASME) Code, Section III, Subsection NC, 1971 Edition, with the
exception of the tees supporting the vacuum breakers.
The downcomer and SRV discharge lines penetrate through the
diaphragm slab and terminate in the suppression pool. A steel ring
plate is welded to the outside of the downcomers. The downcomer/ring
plate assemblies are embedded in the diaphragm slab with the top
surface of the ring plate flush with the drywell side of the
diaphragm slab. All connections are welded to form a continuous
steel membrane between the liner plate and downcomer penetrations.
The SRV discharge lines are routed through welded flued heads at the
diaphragm floor. The flued head design and construction are similar
to the downcomer penetrations and also provide a continuous steel
barrier. The downcomer and SRV discharge lines are designed and
constructed to safety-related requirements. In addition, they are
designed for all postulated loading conditions, including seismic,
hydrodynamic, pressure, and temperature loads. The conservative
design requirements ensure that the SRV discharge and the downcomer
lines will not contribute to bypass leakage.
The diaphragm floor is a reinforced concrete slab approximately
3.5 feet thick. The drywell side surface of the diaphragm slab is
capped with a 1/4 inch thick carbon steel liner plate. The liner
plate and diaphragm slab provide a barrier against the potential for
bypass leakage through the diaphragm floor. The structural integrity
of the diaphragm floor and penetrations was demonstrated during the
pre-operational test program. The drywell was pressurized to a
drywell-to-suppression chamber differential pressure of above 30
psid, which envelopes the maximum drywell-to-suppression chamber
differential pressure postulated to occur during LOCA conditions.
There have been six Unit 1 and three Unit 2 bypass leak tests
performed in accordance with TS Surveillance Requirement 4.6.2.1.d.
These tests were conducted at a drywell-to-suppression chamber
differential pressure of at least 4.0 psid. The measured leakage
area includes leakage from both the vacuum breakers and sources
other than vacuum breakers.
In all cases, the measured leakage is significantly less than
the TS and design values. The maximum measured leakage areas are
0.0400 in2 and 0.0111 in2 for Unit 1 and Unit 2,
respectively; or 5.56% and 1.55 %, respectively, of the TS limit.
The average values are 0.0180 in2 for Unit 1 and 0.0107
in2 for Unit 2; or 2.5% and 1.49%, respectively, of the TS
limit of 0.720 in2. The minimum measured leakage areas are 0.0
in2 and 0.0100 in2 for Unit 1 and Unit 2, respectively, or
0% and 1.3 %, respectively, of the TS limit. Clearly, the test data
confirm that the bypass leakage measured to date at LGS has been
negligible.
In addition, we have obtained bypass leakage data from the
Pennsylvania Power and Light Company, Susquehanna Steam Electric
Station (SSES), Units 1 and 2, which also has Mark II containments
with the Anderson Greenwood vacuum breakers (i.e., the same
manufacturer as the vacuum breakers installed in the LGS, Unit 1 and
Unit 2 containments) and therefore the data is applicable to LGS.
The maximum bypass leakage area for the SSES Unit 1 containment was
0.037 in2, and 0.009 in2 for the SSES Unit 2 containment,
or 4.81% and 1.17%, respectively, of the SSES TS limit. Approval for
a similar TS change for SSES, Units 1 and 2 was issued by the NRC by
letter dated August 11, 1993.
The remaining and most likely source of potential bypass leakage
is the four sets of drywell-to-suppression chamber vacuum breakers.
Each set consists of two vacuum breakers in series, flange mounted
to a tee off the downcomers in the suppression chamber airspace. The
drywell-to-suppression chamber bypass leak test is currently
required by TS Surveillance requirement 4.6.2.1.d to be completed
during each refueling outage and the results are used to verify that
the total bypass area, including that due to the vacuum breakers,
meets the TS limit. If maintenance has been performed on the vacuum
breakers, this test also serves as a post-maintenance vacuum
breakers leakage area test.
The proposed TS changes decrease the frequency of the drywell-
to-suppression chamber bypass leak test. The drywell-to-suppression
chamber bypass leak test data obtained following vacuum breakers
maintenance cannot be utilized to determine vacuum breakers leakage
reliability over the duration of the proposed test interval
extension. To address this concern and collect additional vacuum
breakers leakage data, the proposed TS changes include an additional
requirement to perform a vacuum breaker leakage test as described
below.
The leakage test will be conducted on each set of vacuum
breakers (i.e., four vacuum breakers sets per unit) during each
refueling outage when the drywell-to-suppression chamber bypass leak
test would not be required to be performed. If maintenance is
performed on the vacuum breaker assemblies, this additional test
will be performed post-maintenance to verify that the leakage is
acceptable. This test will be conducted at a drywell-to-suppression
chamber differential pressure of 4.0 psid (i.e., the same as
differential pressure required for the drywell-to-suppression
chamber bypass leak test) by either pressurizing the drywell side of
the vacuum breakers or inducing a vacuum on the suppression chamber
side of the vacuum breakers. The acceptance criteria for the vacuum
breaker leakage tests will be as follows. The total vacuum breaker
leakage areas for all four sets of vacuum breakers will be less than
or equal to 24% of the TS limit (i.e., 0.24 x 0.720 in2 = 0.173
in2). This proposed acceptable vacuum breaker leakage area
provides a 76% margin to the TS limit to account for the leakage
paths other than the vacuum breakers. As described above, previous
bypass leakage testing measured a maximum bypass leakage area of
5.56% of the TS limit. The 76% margin is sufficiently large to
accommodate the other expected leakage sources. In addition, each
set of vacuum breakers will be limited to a leakage area twice the
assumed leakage from a single vacuum breaker set, assuming the
leakage area is evenly distributed among the four sets of vacuum
breakers (i.e., four sets equate to 24% of the TS Limit where each
set is 6% and twice this total is 12% of the TS Limit). This allows
a leakage of less than or equal to 0.0865 in2 (i.e., (0.173
in2 divided by 4 sets of vacuum breakers) x (a factor of 2
times the acceptable total) = 0.0865 in2) for an individual set
of vacuum breakers. This criterion is stipulated to identify
individual sets of vacuum breakers with higher leakage area.
The drywell-to-suppression chamber bypass leak test data
obtained during previous testing at LGS demonstrates conformance by
a large margin compared to the TS and design leakage requirements.
The test data indicates that there is negligible risk that the
bypass leakage will change adversely in future years. Furthermore,
the proposed test frequency is judged to be acceptable based on the
risk of the leakage sources other than the vacuum breakers being
essentially equivalent to that of the rest of the primary
containment structure, which is leak tested (i.e., CILRT) every 40
+/- 10 months as required by TS Surveillance Requirement 4.6.1.2.a.
A bypass leak test will be developed and conducted to verify
acceptable vacuum breaker bypass leakage areas for those outages
when the bypass leak test will not be required to be performed. The
proposed vacuum breaker leakage test with stringent acceptance
criteria, combined with other negligible leakage areas, provide an
acceptable level of assurance that the bypass leakage can be
measured and an adverse condition can be detected and corrected such
that the existing level of confidence that the primary containment
will function as required during a LOCA is maintained.
Therefore, the proposed TS changes will not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes involve the drywell-to-suppression
chamber bypass leak test frequency. There are no physical or
operational changes as a result of these proposed changes. These
proposed changes include the requirement to perform an additional
surveillance test on the vacuum breaker assemblies, applying a
differential pressure of 4.0 psid which is the same differential
pressure as currently required by TS for the drywell-to-suppression
chamber bypass leak test. This required test will ensure that
acceptable vacuum breaker leakage is maintained during those
intervals when the drywell-to-suppression chamber bypass leak test
is not required to be performed. Furthermore, the affected structure
(i.e., primary containment) acts as an accident mitigator and not as
an accident initiator. Accordingly, the possibility of a different
type of malfunction of equipment or the possibility of an accident
of a different type is not introduced.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The drywell-to-suppression chamber bypass leak test data
obtained during previous testing at LGS demonstrates conformance by
a large margin to the TS and design leakage requirements. The test
data indicate that there is negligible risk that the bypass leakage
will change adversely in future years. Furthermore, the proposed
test frequency is judged to be acceptable based on the risk of
sources of leakage other than the vacuum breakers being essentially
equivalent to that of the rest of the primary containment structure,
which is tested every 40 +/- 10 months. A bypass leak test will be
developed and conducted to verify acceptable vacuum breaker bypass
leakage areas for those outages when the bypass leak test will not
be required to be performed. The proposed vacuum breaker leakage
test with stringent acceptance criteria, combined with the other
negligible potential leakage areas, provide an acceptable level of
assurance that the bypass leakage can be measured and an adverse
condition can be detected and corrected such that the existing
levels of confidence that the primary containment will function as
required during a LOCA is maintained.
Therefore, the consequences of an accident are not impacted by
this change and containment integrity during a LOCA will be
maintained.
Therefore, the proposed TS changes do not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Larry E. Nicholson, Acting
Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 17, 1993
Description of amendment request: The proposed Technical
Specification (TS) changes to Surveillance Requirements would eliminate
unnecessary emergency diesel generator (EDG) testing when a diesel
generator or an offsite power source becomes inoperable. The proposed
change would reduce the stresses on the diesel generators caused by
unnecessary testing.
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 because implementation of the proposed TS change, which
would delete the requirement to demonstrate the operability of an
otherwise operable EDG once the potential for a common cause failure
has been dismissed, does not affect the design or performance
characteristics of an EDG. Similarly, deleting the requirement to
demonstrate the operability of EDGs when an offsite power source is
inoperable does not affect the design or performance characteristics
of an EDG. Therefore, the EDGs will maintain their ability to
perform their design function. The EDGs are not assumed to be an
initiator of any analyzed event. The role of the EDGs is the
mitigation of accident consequences. Therefore, this proposed TS
change does not increase the probability of an accident previously
evaluated.
The consequences of an accident previously evaluated could be
affected by the proposed TS change. As described above,
implementation of the proposed change will result in the EDGs
maintaining their ability to perform their design function.
Excessive testing of EDGs can cause reduced reliability. Precluding
unnecessary testing of operable EDGs will improve EDG reliability
and thereby have an overall positive affect on plant safety.
Therefore, this proposed TS change does not increase the
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
because implementation of the proposed TS change will not involve
physical changes to plant systems, structures, or components (SSC).
The design or performance characteristics of the EDG will not be
affected by the proposed change. The proposed change does not
introduce any new modes of plant operation or make any changes to
system setpoints which would initiate a new or different kind of
accident. Therefore, the possibility of a new or different kind 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 because the proposed TS change does not affect
the design or performance of any EDG. The change will increase EDG
reliability by reducing the stresses on the EDG from unnecessary
testing. This will result in an overall increase in plant safety.
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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 19, 1993
Description of amendment request: The proposed change would
eliminate the listing of specific position titles for the Plant
Operations Review Committee (PORC) composition in favor of allowing the
Plant Manager to appoint PORC members. This would eliminate the need to
change the Technical Specifications (TSs) in the future whenever a
position title is changed.
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 because the proposed TS change is administrative in
nature. The PORC member titles will be removed from the TS to
facilitate not requiring that a TS change be submitted for NRC
approval when position titles change. PORC member qualifications
will continue to be consistent with those required for the Facility
Staff and meet or exceed Sections 4.2, 4.4, or 4.6 of ANSI N18.1-
1971. The proposed change ensures that PORC will continue to be
comprised of personnel involved in daily plant activities who are
experienced individuals with varied expertise. By maintaining the
qualification requirements for members of PORC who represent various
areas of expertise, PORC will continue to fulfill its requirements
specified in TS Section 6.5.1.6. The proposed change does not
involve any physical changes to plant systems, structures, or
components (SSC), or the manner in which these SSC are operated,
maintained, modified, tested, or inspected. Therefore, the proposed
TS 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
because implementation of the proposed TS change will not involve
physical changes to plant SSC or the manner in which these SSC are
operated, maintained, modified, tested or inspected. The proposed
change does not introduce any new modes of plant operation or make
any changes to system setpoints which would initiate a new or
different kind of accident. Therefore, the possibility of a new or
different kind 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 because the proposed TS change is
administrative in nature by providing internal flexibility in
changing organizational titles and does not reduce the PORC function
or responsibilities. PORC will continue to be filled by
appropriately qualified personnel who have a variety of expertise.
The change does not affect the plant material condition, operation,
or accident analyses. 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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Charles L. Miller
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: November 24, 1993
Description of amendments request: The proposed changes to the
Technical Specification will relocate the reactor trip system and
engineered safety feature actuation system response time limits from
the TS to the Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature and do not involve
any change to the configuration or method of operation of any plant
equipment used to mitigate the consequences of an accident. Also,
the proposed changes do not alter the conditions or assumptions in
any of the FSAR accident analyses. Since the FSAR accident analyses
remain bounding, the radiological consequences previously evaluated
are not adversely affected by the proposed changes. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed changes are administrative in nature and do
not involve any change to the configuration or method of operation
of any plant equipment used to mitigate the consequences of an
accident.Accordingly, no new failure modes have been defined for any
plant system or component important to safety nor has any new
limiting failure been identified as a result of the proposed
changes. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety. The proposed changes are administrative in
nature and will continue to ensure that the response times for the
RTS and ESFAS instrumentation do not exceed the limits assumed in
the accident analyses. As a result of the proposed changes, response
time limits for the RTS and ESFAS will be administratively
controlled in accordance with the provisions of 10 CFR 50.59, thus
eliminating an unnecessary burden of governmental regulation without
reducing protection for public health and safety. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: James H. Miller, III, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: S. Singh Bajwa
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and
50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3,
Limestone County, Alabama
Date of amendment request: September 30, 1993 (TS 337)
Description of amendment request: The proposed amendments provide
an administrative vehicle for modifying a condition of the facility
operating license for each of the BFN units. The condition requires the
licensee to implement and maintain in effect all provisions of the
``Fire Protection Program (FPP)'' and lists the U.S. Nuclear Regulatory
Commission (NRC) staff safety evaluations (SE) approving the FPP. If
the staff approves of a revision currently under review to an element
of the FPP, the ``Appendix R Safe Shutdown Program (SSP)'', the
proposed amendments would add the staff SE documenting approval of the
revised SSP to the above listing of SEs in each facility operating
license. The current SSP is directed toward the safe shutdown of only
one operating plant (Unit 2). The revised SSP would be directed toward
the safe shutdown of two operating plants (Units 2 and 3).
Additionally, the proposed amendments add the definition of the SSP
to Section 1.0 of the Unit 3 Technical Specifications (TS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This proposed change is administrative in nature. The proposed
change is being made to revise the license condition to reflect a
combined Unit 2 and 3 Appendix R Safe Shutdown Program following NRC
approval. Compliance with the applicable Appendix R requirements is
ensured through implementation of the Fire Protection Program and
the Appendix R Safe Shutdown Program. The change does not affect any
design bases accident or the ability of any safe shutdown equipment
to perform its function. Also, there are no physical modifications
required to implement this TS change. Therefore, these proposed
administrative changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change is administrative in nature. The proposed
change is being made to revise the license condition to reflect a
combined Unit 2 and 3 Safe Shutdown Program following NRC approval.
Compliance with the applicable Appendix R requirements is ensured
through implementation of the Fire Protection Program and Appendix R
Safe Shutdown Program. This change does not affect any design basis
accident or the ability of any safe shutdown equipment to perform
its function. Also, there are no physical modifications required to
implement this TS change. Therefore, these proposed administrative
changes do not create the possibility of a new or different kind of
accident from an accident previously evaluated.
3. This change does not involve a significant reduction in the
margin of safety.
The proposed changes are administrative in nature. Compliance
with the applicable Appendix R requirements is ensured through the
implementation of the Fire Protection Program and Appendix R Safe
Shutdown Program. The proposed change does not affect any design
basis accident and does not reduce or adversely affect the
capability to achieve and maintain safe shutdown in the event of a
fire. Furthermore, no reductions to the requirements for equipment
operability, surveillance requirements or setpoints are being made
which could result in reduction in the margin of safety. Therefore,
these proposed administrative changes will not result in a reduction
in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Mr. Frederick J. Hebdon
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and
50-296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3,
Limestone County, Alabama
Date of amendment request: October 12, 1993 (TS 320)
Description of amendment request: The proposed amendment would
delete reference in the BFN Unit 3 Technical Specifications (TS) to the
Reactor Water Cleanup (RWCU) system floor drain high temperature
switches and the RWCU system space high temperature switches. The
piping configuration for the Unit 3 RWCU system has been modified, and
the licensee contends that its revised High Energy Line Break (HELB)
analysis has demonstrated that these switches are no longer required.
Instead, to initiate RWCU system isolation, the HELB analysis has
indicated the need for temperature switches in the main steam vault,
the heat exchanger room, and the RWCU pipe trench. The proposed
amendment therefore would add temperature switches to the Unit 3 TS for
these areas and modify the TS Bases section accordingly. The proposed
amendment also adds clarifying remarks to Tables 3.2.A and 3.2.B of the
TS for each of the BFN units. The proposed remarks list the actuation
signals for the various primary containment valve group isolations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
An analysis of HELBs in the Unit 3 reactor building identified
certain RWCU pipe breaks which could not be automatically detected
and isolated in a reasonable time frame. To resolve this issue, a
design change is being performed to remove from service the existing
non-environmentally qualified temperature switches used to detect
RWCU line breaks and replace them with environmentally qualified
RTDs [resistance temperature detectors] and IEEE [Institute of
Electrical and Electronics Engineers] Class 1E qualified ATUs
[analog trip units] located to detect and isolate the critical RWCU
pipe breaks. This TS amendment adds the new ATUs [sic] function to
Tables 3.2.A and 4.2.A. Note 14 is deleted from Table 3.2.A since it
only applies to the temperature switches being removed from the
table.
The safety function of the RTD/ATU temperature loops is to
provide an isolation signal to close the RWCU suction line isolation
valves (FCV-69-001 and FCV-69-002) and RWCU return line valve (FCV-
69-012) on a high area temperature. This ensures RWCU pipe breaks
are isolated. No other RWCU safety functions are affected by the
change.
The new RTD/ATU temperature loops were chosen to decrease the
time required to initiate closure of the RWCU valves. This improves
the detection/isolation of RWCU breaks and helps to limit the
reactor coolant lost, helps ensure core cooling, and helps ensure
that environmental conditions inside the reactor building are
maintained within the required limits.
Components added by this change are qualified for the
environment in which they will operate. This ensures that the system
will perform its function in a post accident environment. No
additional paths for the release of radiation or contamination are
created. The failure modes of the RTDs and ATUs are such that any
single failure will result in a gross failure alarm and/or a channel
trip. Because of the redundancy, separations, and logic designed
into the system, a single failure of any part of the system will not
prevent isolation of the primary containment isolation valves and
spurious operation is minimized. The RTDs will be located and the
instrument setpoints will be set to preclude spurious trips due to
ambient temperatures including localized hot areas while assuring a
timely trip due to a pipe break. Therefore, the proposed amendment
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This change is being made to improve the RWCU leak detection/
isolation function of the RWCU Primary Containment Isolation System
(PCIS). The PCIS will perform its intended safety function in the
same manner as the previous installation. There is no affect [sic]
on the function or operation of any other plant system.
Failure of the RTD/ATU temperature loops would be no different
than failure of existing temperature switches. Since environmental
qualification requirements, divisional separation, single failure
requirements and one-out-of-two taken twice logic requirements are
maintained, the possibility of a RWCU isolation failure on a RWCU
line break or of a spurious isolation is no more likely after the
change than before.
In the existing design, logic relays are powered from RPS Bus A
or B. The new design also uses RPS Bus A or B to feed the ATUs.
Therefore, the consequence of a power failure is unchanged from the
present design. The seismic qualification and proper circuit
coordination of the installation is maintained. The system functions
and operates in the same manner as previously evaluated in the
Safety Analysis Report. No new system interactions other than
additional RTDs located in the main steam valve vault to input into
the PCIS logic for isolation of the RWCU have been introduced by
this activity. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The margin of safety will be enhanced by installing instruments
that provide quicker response to a temperature rise indicative of a
pipe break. Calculations have been performed to determine the
analytical limits for the RTD/ATU temperature loops in each of the
monitored areas and to determine the setpoints for the ATUs in each
area. The setpoints are set above the maximum expected room
temperatures to avoid spurious actuations due to ambient conditions
and below the analytical limits to ensure timely detection of a pipe
break. This type of design utilizing ATUs has been analyzed by the
NRC [U.S. Nuclear Regulatory Commission staff] (NEDO-21617, Analog
Transmitter/Trip Unit System for Engineered Safeguard Sensor Trip
Input) and has been found to be generically acceptable at BWR
facilities. Therefore, the proposed amendment does not involve a
significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Mr. Frederick J. Hebdon
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: December 10, 1993
Description of amendment request: The proposed change would change
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2).
Specifically, the proposed changes would modify the surveillance
frequency of the Auxiliary Feedwater System pumps from monthly to
quarterly in accordance with the guidance provided in Generic Letter
93-05, ``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,'' dated
September 27, 1993.
The NRC has completed a comprehensive examination of surveillance
requirements in TS that require testing at power. The evaluation is
documented in NUREG-1366, ``Improvements to Technical Specification
Surveillance Requirements,'' dated December 1992. The NRC staff found,
that while the majority of testing at power is important, safety can be
improved, equipment degradation decreased, and an unnecessary burden on
personnel resources eliminated by reducing the amount of testing at
power that is required by TS. Based on the results of the evaluations
documented in NUREG 1366, the NRC issued Generic Letter 93-05.
The Auxiliary Feedwater System supplies water to the steam
generators to remove decay heat from the Reactor Coolant System. To
ensure operability of the Auxiliary Feedwater System, the pumps are
currently tested on a monthly basis as required by the TS. Consistent
with Generic Letter 93-05, Item 9.1 and NUREG-1366, the licensee is
requesting a change to the surveillance testing frequency for the
Auxiliary Feedwater Pumps from monthly to quarterly on a staggered test
basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of North Anna Power Station in
accordance with the proposed Technical Specifications changes will
not:
1. Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
Changing the surveillance test frequencies of the Auxiliary
Feedwater System pumps does not significantly affect the probability
of occurrence or consequences of any previously evaluated accidents.
Quarterly testing of the pumps on a staggered basis will continue to
assure that the Auxiliary Feedwater System will be capable of
performing its intended functions. Therefore, the change in
frequency of testing the Auxiliary Feedwater System pumps does not
affect the probability or consequences of any previously analyzed
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Changing the surveillance test frequency of the Auxiliary
Feedwater System pumps does not involve any physical modification of
the plant or result in a change in a method of operation. Quarterly
testing of the Auxiliary Feedwater System pumps on a staggered basis
will continue to assure that the Auxiliary Feedwater System will be
capable of performing its intended function. Therefore, a new or
different type of accident is not made possible.
3. Involve a significant reduction in a margin of safety.
Changing the surveillance test frequency of the Auxiliary
Feedwater System pumps does not affect any safety limits or limiting
safety system settings. System operating parameters are unaffected.
The availability of equipment required to mitigate or assess the
consequence of an accident is not reduced. Quarterly testing of the
Auxiliary Feedwater System pumps on a staggered basis will continue
to assure that the Auxiliary Feedwater System will be capable of
performing its intended functions. Safety margins are, therefore,
not decreased.
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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow
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 20555, and at the local public document
rooms for the particular facilities involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: October 5, 1993
Brief description of amendment: The proposed change to the
Technical Specifications would revise the wording of liquid release
rate limit and its associated bases, and relocate the old 10 CFR 20.106
requirements to the new 10 CFR 20.1302 to be consistent with the
revised terminology of 10 CFR Part 20. The new wording will retain the
same overall level of effluent control required to meet the design
objectives of Appendix I to 10 CFR Part 50.
Date of issuance: December 14, 1993
Effective date: December 14, 1993
Amendment No.: 40
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59746) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 14, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: August 5, 1992
Brief description of amendments: The amendment revises the
Braidwood Station, Units 1 and 2, Technical Specifications (TS)
regarding Engineered Safety Features Actuation System (ESFAS)
instrumentation. The ESFAS, Functional Units, Analog Channel
Operational Test interval is changed from monthly to quarterly.
Eighteen changes to the Reactor Trip System (RTS) are also included in
this TS change.
Date of issuance: December 16, 1993
Effective date: December 16, 1993
Amendment Nos.: 44 and 44
Facility Operating License Nos. NPF-72 and NPF-77. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 28, 1992 (57 FR
48816) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 16, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: October 29, 1993
Brief description of amendments: The amendments revise Table 3.6.3-
1, ``Primary Containment Isolation Valves,'' of the LaSalle Technical
Specifications for Units 1 and 2 by adding a new category of valves to
these tables. There are a total of eight new valves added in each
table, consisting of two check valves in each of four backfill lines.
The backfill lines were added in response to NRC Bulletin 93-03,
``Resolution of Issues Related to Reactor Vessel Water Level
Instrumentation in BWRs,'' dated May 28, 1993.
Date of issuance: December 10, 1993
Effective date: December 10, 1993
Amendment Nos.: 92 and 76
Facility Operating License Nos. NPF-11 and NPF-18. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1993 (58 FR
59493). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 10, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: November 25, 1992, as
supplemented by letter dated February 5, 1993.
Brief description of amendment: The amendment revises surveillance
intervals for Process Radiation Monitors, Area Radiation Monitors, the
Main Steam Line Radiation Monitors, the Auxiliary Feedwater System
Initiating Logic, the Main Steam Safety Valves Setpoints, and the Toxic
Gas Detection System Monitors to accommodate a 24-month refueling
cycle. These revisions are being made in accordance with the guidance
provided by Generic Letter 91-04, ``Changes in Technical Specification
Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''
Date of issuance: December 16, 1993
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 166
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 25, 1993 (58 FR
16219) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 16, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 5, 1993, as
supplemented November 15 and 22, 1993
Brief description of amendments: The amendments revise the
Technical Specifications to reflect the appropriate operability
requirements for cold leg accumulator water volume and surveillance
requirements values for the centrifugal changing pumps, safety
injection pumps, and residual heat removal pumps to prevent possible
runout conditions during a loss of coolant accident event.
Date of issuance: December 15, 1993
Effective date: December 15, 1993
Amendment Nos.: 110 and 104
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57848) The November 15 and 22, 1993, letters provided clarifying
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 December 15, 1993. No significant hazards
consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 5 and 14, 1993, as
supplemented November 15 and December 14, 1993
Brief description of amendments: The amendments revise the
Technical Specifications to allow the implementation of interim steam
generator tube plugging criteria for the tube support plate elevations.
Date of issuance: December 16, 1993
Effective date: December 16, 1993
Amendment Nos.: 111 and 105
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57849) The November 15 and December 14, 1993, letters provided
clarifying information and revisions to the coolant specific activity
that did not change the scope of the original application and 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 December 16, 1993. No significant hazards
consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 7, 1993
Brief description of amendments: The amendments revise the
Technical Specification to (a) reduce the slope of the axial power
imbalance penalty in the overtemperature-delta temperature reactor
protection system trip setpoint equation, and (b) increase the boron
concentration limits in the cold leg accumulators, the refueling water
storage tank, the reactor coolant system, and refueling canal during
MODE 6 conditions. These changes reflect the reloading of Unit 1 with
Mark BW fuel for Cycle 8 including an increase in cycle length from 350
effective full power days (EFPD) to 390 EFPD.
Date of issuance: December 17, 1993
Effective date: December 17, 1993
Amendment Nos.: 112 and 106
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57847) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 17, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: October 25, 1993, as
supplemented December 3 and 6, 1993
Brief description of amendments: The amendments revise the
Technical Specifications (TS) Figure 2.1-1, certain TS Table 2.2-1
factors in the equation for the OVERTEMPERATURE delta T and OVERPOWER
delta T setpoints, and Figure 3.2-1 to reflect a reduction in the
required minimum measured reactor coolant system (RCS) flow rate from
385,000 gallons per minute (gpm) to 382,000 gpm for Unit 1. Catawba
Unit 2 values are unchanged and, accordingly, certain TS pages were
modified to retain the current TS values in effect for Unit 2.
The need for these changes is attributed to the effects of steam
generator tube plugging and to a hot leg temperature streaming
phenomenon. The application also proposed to revise the text of TS
2.1.1 and the definition for TS Figure 2.1-1. These changes are not
related to the changes in RCS flow rate. The staff is continuing to
review these proposed changes and, accordingly, they are not dealt with
in this amendment.
Date of issuance: December 17, 1993
Effective date: Effective within 30 days of its date of issuance.
Amendment Nos.: 113 and 107
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59747) The December 3 and 6, 1993, letters provided clarifying
information that did not change the scope of the October 25, 1993,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 17, 1993. No significant hazards
consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: May 20, 1993
Brief description of amendment: The amendment removed unnecessary
operability requirements for the Intermediate Range Monitors (IRMs) and
the Average Power Range Monitors (APRMs) during plant shutdown
operations.
Date of issuance: December 13, 1993
Effective date: December 13, 1993
Amendment No: 109
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34077) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 13, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, Post Office Box 1406, S. Commerce at Washington, Natchez,
Mississippi 39120.
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: July 23, 1993
Brief description of amendments: The amendments are necessary to
implement new Standards for Protection Against Radiation (10 CFR Part
20).
Date of issuance: December 16, 1993
Effective date: December 16, 1993
Amendment Nos.: 125, 63
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46234) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 16, 1993. 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
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: October 8, 1993
Brief description of amendment: The amendment deletes portions of
the Oyster Creek Nuclear Generating Station Radiological Effluent
Technical Specifications and relocates them to controlled programs in
accordance with the guidance contained in NRC Generic Letter 89-01,
dated January 31, 1989.
Date of issuance: December 13, 1993
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 166
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59749) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated December 13, 1993. No
significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: October 18, 1993
Brief description of amendment: The amendment revises the Technical
Specifications to delete requirements to demonstrate by testing, that a
redundant system/component is operable when a system/component is
declared inoperable. In lieu of testing the redundant system/component
to demonstrate its operability the Technical Specifications are being
revised to require an administrative check of plant records to verify
operability of the redundant system/component. Confirming changes are
made to Definition 1.1 ``Operable-Operability.''
Date of issuance: December 21, 1993
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 167
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59749) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 21, 1993. No
significant hazards consideration comments received: Yes. Comments were
provided by letter dated December 10, 1993, from the State of New
Jersey, Department of Environmental Protection and Energy, Division of
Environmental Safety, Health and Analytical Programs. The comments and
the NRC staff's response are addressed in the Commission's Safety
Evaluation dated December 21, 1993.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: September 20, 1993, as
supplemented on October 1, 1993.
Brief description of amendment: The amendment revises the plant
Technical Specifications to reflect a partial GPU Nuclear
reorganization to become effective when Three Mile Island, Unit 2 (TMI-
2), enters the Post-Defueling Monitored Storage (PDMS) mode. This
reorganization includes deleting TMI-2 as a Division and incorporating
those functions and responsibilities required to maintain the PDMS
condition and requirements into the current TMI-1 Division. The TMI-1
Division will be renamed the TMI Division.
Date of issuance: December 13, 1993
Effective date: No specific date has been specified by the staff
for the effectiveness of this amendment. The amendment will become
fully effective at such time as the Vice President - TMI has been
delegated the full responsibility of the overall safe operation of both
TMI-1 and TMI-2.
Amendment No.: 179
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52987). The October 1, 1993, submittal provided clarifying and
corrected TS pages which did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of this amendment is contained in a Safety
Evaluation dated December 13, 1993. No significant hazards
consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: August 9, 1993
Brief description of amendment: The amendment revises the plant
Technical Specifications to be consistent with a major revision to 10
CFR Part 20 that is to be implemented by January 1, 1994.
Date of issuance: December 21, 1993
Effective date: As of the date of issuance to be implemented on
January 1, 1994.
Amendment No.: 180
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59751). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 21, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: August 24, 1993.
Brief description of amendment: The amendment revises the plant
Technical Specifications to adopt the Standard Technical Specification
(STS) provision that allows a period up to 24 hours to complete a
surveillance requirement upon the discovery that the surveillance has
been missed.
Date of issuance: December 22, 1993
Effective date: As of its date of issuance, to be implemented
within 30 days of issuance.
Amendment No.: 181
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59751). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 22, 1993. No
significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
Gulf States Utilities Company and Cajun Electric Power Cooperative,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana
Parish, Louisiana
Date of amendment request: January 13, 1993, as supplemented by
letter dated October 18, 1993.
Brief description of amendment: The amendment revises the River
Bend, Unit 1 operating license to reflect a change in ownership of Gulf
States Utilities (GSU). GSU, which owns a 70 percent undivided interest
in the River Bend Station, will become a wholly-owned subsidiary
company of Entergy Corporation.
Date of issuance: December 16, 1993
Effective date: December 6, 1993, to be implemented within 180 days
of issuance.
Amendment No.: Amendment No. 69
Facility Operating License No. NPF-47: The amendment revised the
license.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36435) The October 18, 1993, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 16, 1993. No significant hazards consideration comments
received: Yes. Comments and a request for hearing were received from
Cajun Electric Power Cooperative of Baton Rouge, Louisiana.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
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: January 13, 1993, as supplemented by
letter dated June 29, 1993.
Brief description of amendment: The amendment revises the River
Bend Station, Unit 1 operating license to include as a licensee,
Entergy Operations, Inc. (EOI), and to authorize EOI to use and operate
River Bend and to possess and use related licensed nuclear materials.
Date of issuance: December 16, 1993
Effective date: December 16, 1993 to be implemented within 180 days
of issuance.
Amendment No.: 70
Facility Operating License No. NPF-47: The amendment revised the
license.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36436) The June 29, 1993, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 16, 1993. No significant hazards consideration comments
received: Yes. Comments and a request for hearing were received from
Cajun Electric Power Cooperative of Baton Rouge, Louisiana.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: June 18, 1993
Brief description of amendment: The proposed changes to Technical
Specifications 6.2.3.1, ``Independent Safety Engineering Group (ISEG)
Function;'' 6.2.3.4, ``ISEG Records;'' 6.4.1, ``Training;'' and
6.5.2.2, ``Nuclear Review and Audit Group (NRAG) Composition'' are
editorial changes reflecting recent administrative/organizational
changes which occurred at Clinton Power Station.
Date of issuance: November 29, 1993
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 86
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43927) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 29, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library District, 310 N. Quincy Street, Clinton, Illinois 61727
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of application for amendment: April 16, 1993, as supplemented
September 28 and December 3, 1993
Brief description of amendment: The amendment revises Technical
Specifications to allow certain tests normally designated as 18-month
surveillances to be delayed until the next refueling outage scheduled
to begin August 6, 1994.
Date of issuance: December 22, 1993
Effective date: December 22, 1993
Amendment No.: 158
Facility Operating License No. DPR-74. Amendments revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41505) The supplemental letters provided clarifying information which
did not change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated December 22, 1993.
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. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: June 7, 1993
Brief description of amendment: The amendment revises Technical
Specification (TS) 3/4.8.1, ``AC Sources-Operating,'' and associated
Bases to eliminate unnecessary diesel generator testing when a diesel
generator or an offsite power source becomes inoperable. The amendment
is intended to increase diesel generator reliability and the overall
level of plant safety by reducing the stresses on the diesel generators
caused by unnecessary testing. The amendment also makes additional
changes to TS 3/4.8.1 to further enhance diesel generator reliability
and incorporate certain administrative changes.
Date of issuance: December 15, 1993
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 54
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36440) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 15, 1993. 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.
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: November 11, 1992, as supplemented by
letters dated July 2, 1993, and November 24, 1993.
Description of amendment request: The amendment modifies the
Seabrook Station Technical Specifications to allow the use of either
the fixed incore detector system or the movable incore detector system
to perform technical specification surveillances. Specifically, the
amendment modifies Technical Specification sections 3.1.3, 4.2.2,
4.2.3, 4.2.4, and 3.3.3.
Date of issuance: December 22, 1993
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 27
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 3, 1993 (58 FR
7002). The licensee's letters dated July 2, 1993, and November 24,
1993, provide additional information and clarification to the
application but do not change the initial proposed no significant
hazards consideration determination and do not provide information
outside the scope of the original Federal Register notice. The
licensee's November 24, 1993, letter provides a commitment to acquire,
through the end of Cycle 4, a limited number of flux maps using the
movable incore detector system for comparison to flux maps obtained
using the fixed incore detector system. Additionally, the licensee
committed to provide a report to the NRC at the end of Cycle 4
regarding comparison of the data obtained from both systems. The NRC's
approval of the requested TS changes is conditioned upon the licensee's
implementing the commitment. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated December 22, 1993.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
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 19, 1993
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to reflect staff positions and improvements to the
TS in response to Generic Letter 90-06, ``Resolution of Generic Issue
70, `Power-Operated Relief Valve and Block Valve Reliability, and
Generic Issue 94, `Additional Low-Temperature Overpressure Protection
for Light Water Reactors.''' Generic Issue 94 was closed out by
Amendment 80 dated July 12, 1993. With the issuance of this TS
amendment, we consider the licensee's response to Generic Letter 90-06
and Generic Issue 70 (TAC No. M77362) complete for the Millstone
Nuclear Power Station, Unit No. 3.
Date of issuance: December 16, 1993
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 88
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 9, 1993 (58 FR
32388) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 16, 1993. No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Pennsylvania Power and Light Company, Docket No. 50-387,
Susquehanna Steam Electric Station, Unit 1, Luzerne County,
Pennsylvania
Date of application for amendment: July 21, 1993
Brief description of amendment: The amendment revised the Technical
Specifications to modify the requirement for acquisition of baseline
data on single-loop operation from during startup testing following
each refueling outage to at least once per 18 months.
Date of issuance: December 10, 1993
Effective date: December 10, 1993
Amendment No.: 131
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41509) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 10, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Portland General Electric Company, et al., Docket No. 50-344,
Trojan Nuclear Plant, Columbia County, Oregon
Date of application for amendment: April 1, 1993 as supplemented
June 9 and August 5, 1993.
Brief description of amendment: This amendment relocates the
Radiological Effluent Technical Specifications (RETS) to the Offsite
Dose Calculation Manual (ODCM) and Process Control Program (PCP) in
accordance with NRC staff Generic Letter 89-01, and changes the
required frequency for submittal of the radioactive Effluent Release
Report from semiannual to annual in accordance with 10 CFR 50.36(a).
Date of issuance: December 6, 1993
Effective date: 30 days from date of issuance
Amendment No.: 193
Facility Operating License No. NPF-1: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36442) The supplements proposed additional changes and clarification to
the TS regarding the frequency of effluent reporting. The changes were
within the scope of the action described in the notice and did not
change the initial no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 6, 1993. No significant hazards
consideration comments received: No.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207
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: September 25, 1992
Brief description of amendment: The amendment to the Technical
Specifications (TSs) deletes the surveillance requirements for the
iodine analyzer portion of the drywell atmosphere Continuous Atmosphere
Monitoring system from TS Table 4.6-2 and makes accompanying changes to
TS Bases Section 3.6/4.6.D. These changes are consistent with the
guidance in Regulatory Guide 1.45, ``Reactor Coolant Boundary Leakage
Detection Systems.''
Date of issuance: December 9, 1993
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 200
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 25, 1993 (58 FR
16229) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 9, 1993. 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 Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: May 26, 1992
Brief description of amendments: These amendments increase the
shutdown margin requirements for the current operating cycle at both
units; reduce the containment pressure high-high setpoint and allowable
value; and change the containment spray system, containment fan cooler
and service water system response times. These changes were
necessitated by the discovery of containment fan cooler unit and
containment spray system response times greater than originally assumed
for Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB)
analysis, and auxiliary feedwater system flow greater than assumed for
the MSLB analysis.
Date of issuance: December 16, 1993
Effective date: December 16, 1993
Amendment Nos. 149 and 127
Facility Operating License Nos. DPR-70 and DPR-75. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 19, 1992 (57 FR
37571) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 16, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: June 28, 1991
Brief description of amendment: The amendment revised Technical
Specification 3.5.1 to add a required action to periodically monitor
alternative indication if one or both automatic depressurization system
(ADS) safety related instrument air header(s) low pressure alarm system
instrumentation channels become inoperable.
Date of issuance: December 13, 1993
Effective date: December 13, 1993
Amendment No.: 52
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 7, 1991 (56 FR
37590) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 13, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: March 19, 1991
Brief description of amendment: The amendment revised Technical
Specification Table 3.3.3-1 to make the required actions for the
automatic depressurization system (ADS) consistent with the as-built
configuration of the system. An editorial change to Action statement 32
was added to this amendment to make Action statement 32 consistent with
Action statements 30 and 33.
Date of issuance: December 17, 1993
Effective date: December 17, 1993
Amendment No.: 53
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 15, 1991 (56 FR
22480) Additional clarifying information was provided verbally by the
utility on November 5, 1993, that did not change the initial proposed
no significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated December 17, 1993. No significant hazards consideration comments
received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: November 9, 1992 as supplemented
on November 22, 1993
Brief description of amendment: The amendment revises the Technical
Specifications to allow the de-energization of the borated water
storage tank outlet isolation valves in the open position during
operational Modes 1, 2, 3, and 4.
Date of issuance: December 16, 1993
Effective date: December 16, 1993
Amendment No. 182
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28061) The supplemental letter provided additional information that did
not change the proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a safety evaluation dated December 16, 1993. No
significant hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 1, 1991
Brief description of amendment: The amendment revises the Technical
Specification Sections 3.3.3.6, 3.6.4.1, 4.11.2.5, 6.2.2, and Tables
3.3-4 and 3.3-10 to correct typographical errors and make editorial
changes.
Date of issuance: December 21, 1993
Effective date: December 21, 1993
Amendment No.: 86
Facility Operating License No. NPF-30. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 18, 1991 (56
FR 47244) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 21, 1993. No
significant hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity For a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By February 4, 1994, 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 20555 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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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 20555, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Commonwealth Edison Company, Docket No. STN 50-456, Braidwood
Station, Unit 1, Will County, Illinois
Date of application for amendment: November 12, 1993, as
supplemented by letters dated November 18 and December 6, 1993
Brief description of amendment: The amendment changes the existing
technical specifications (TS) by adding a footnote to TS 4.4.5.0 to
address steam generator (SG) operability requirements. The change
references an unscheduled inspection of the 1C SG which occurred due to
a tube leak in that SG. The amendment was required because the
circumstances of the inspection were not covered by the existing TS 3/
4.4.5. It will allow SG operability requirements to be satisfied until
the next SG inservice inspection, scheduled to begin no later than
March 5, 1993.
Date of issuance: December 16, 1993
Effective date: December 16, 1993
Amendment No.: 43
Facility Operating License No. NPF-72. The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated December 16, 1993.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
NRC Project Director: James E. Dyer
Dated at Rockville, Maryland, this 28th of December 1993.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV/V, Office of
Nuclear Reactor Regulation
[Doc. 94-53 Filed 1-4-94; 8:45 am]
BILLING CODE 7590-01-F