[Federal Register Volume 62, Number 19 (Wednesday, January 29, 1997)]
[Notices]
[Pages 4341-4360]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10129]
-----------------------------------------------------------------------
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 January 4, 1997, through January 16, 1997.
The last biweekly notice was published on January 15, 1997 (62 FR
2185).
Notice of Consideration of Issuance of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By February 28, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
[[Page 4342]]
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. 2
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
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 amendment request: December 30, 1996
Description of amendment request: The amendment revises (1)
chemistry data (nickel content) shown on Technical Specification (TS)
Figures 3.4-2 and 3.4-3 for TS 3/4.4.9, ``Pressure/Temperature
Limits,'' and (2) the associated Bases 3/4.4.9 to reflect changes to
chemistry and material properties and changes to comply with recent
U.S. Nuclear Regulatory Commission (NRC) rule changes to 10 CFR 50,
Appendix G.
Basis for proposed no significant hazards 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:
This change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
There are no physical changes to any plant equipment created by
the proposed changes. The chemistry and material property changes do
not impact the ability of the reactor vessel to maintain [its]
pressure boundary integrity as previously evaluated. The decrease in
EOL USE [End-of-Life Upper Shelf Energy] for weld heat 5P6771 is
relatively minor and remains above the required value that has been
prescribed by the NRC to provide the necessary level of ductility
assumed for reactor vessel integrity evaluations. Therefore, the
accident initiating and mitigating aspects of the pressure vessel
are not affected. In addition, neither the proposed change requiring
the ISLH [In-Service Leak and Hydrostatic] test to be complete
before the core is critical nor the proposed change allowing fuel in
the reactor vessel during ISLH affects any accident initiating
mechanisms. The proposed change requiring the ISLH test to be
completed before the core is critical will not increase the
consequences of previously evaluated accidents because it
conservatively assures the core is subcritical. Although the
proposed change allows fuel in the vessel during ISLH utilizing the
ISLH Pressure-Temperature (P-T) limits, the consequences of a
pressure boundary leak have not changed because ISLH testing is
already allowed using the normal plant P-T limits. In addition, the
ISLH will be required to be completed before the core is allowed to
go critical. The consequences of a leak with fuel in the vessel
during ISLH are the same using either the normal P-T limits or the
ISLH limits.
Therefore, there would be no 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.
There are no physical changes to any plant equipment or new
components created by the proposed changes. The chemistry and
material property changes do not impact the pressure boundary
integrity of the reactor vessel. The decrease in EOL USE for weld
heat 5P6771 is relatively minor and remains above the required value
that has been prescribed by the NRC to provide the necessary level
of ductility assumed for reactor vessel integrity evaluations.
Therefore, the accident initiating aspects of the pressure vessel
are not affected. In addition, neither the proposed change requiring
the ISLH test to be complete before the core is critical nor the
proposed change allowing fuel in the reactor vessel during ISLH
creates any new accident initiating mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
[[Page 4343]]
The changes in chemical and material properties do not adversely
affect any reactor vessel integrity evaluations, such as PTS
[Pressurized Thermal Shock] or P-T limits. The USE for weld heat
5P6771 does decrease slightly as described in TS Bases Table B 3/
4.4-1. However, the predicted EOL USE remains above the value
prescribed in 10 CFR 50, Appendix G and is not a significant
reduction in the margin of safety. With regard to the proposed
changes allowing fuel in the reactor vessel during ISLH, the
existing TS Bases specifically state that fuel is not to be in the
reactor vessel when the ISLH P-T curve is utilized. However, this
change is consistent with the revised 10 CFR 50, Appendix G rule and
as such, is not a significant reduction in the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Mark Reinhart, Acting
Commonwealth Edison Company, Docket No. 50-010, Dresden Nuclear
Generating Station, Unit 1, Grundy County, Illinois
Date of amendment request: October 23, 1996
Description of amendment request: The proposed change would amend
the Dresden Unit 1 Appendix A Technical Specifications (TS). The
proposed amendment is a complete revision of the TS to the same format
as Dresden Unit 2/3 TS.
Basis for proposed no significant hazards 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. Will operation of the facility according to this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. In general the proposed amendment represents the conversion
of current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis
(Decommissioning Plan). Implementation of these changes will not
reduce reliability of equipment assumed to operate in the current
safety analysis (Decommissioning Plan), or will provide continued
assurance that specified parameters remain within their acceptance
limits, and as such, will not significantly increase the probability
or consequences of a previously evaluated accident.
Some of the proposed changes represent minor curtailments of the
current requirements which are based on generic guidance or
previously approved provisions for other stations. The proposed
amendment for Dresden Station Unit 1's Technical Specifications in
general is based on STS [Standard Technical Specifications]
guidelines or NRC accepted changes to other facilities such as
Trojan or San Onofre Unit 1. Any deviations from STS requirements do
not significantly increase the probability or consequences of any
previously evaluated accidents for Dresden Station Unit 1. The
proposed amendment is consistent with the current safety analysis
(Decommissioning Plan) and has been previously determined to
represent sufficient requirements for the assurance and reliability
of equipment assumed to operate in the safety analysis
(Decommissioning Plan), or provide continued assurance that
specified parameters remain within their acceptance limits. As such,
these changes will not significantly increase the probability or
consequences of a previously evaluated accident.
2. Will operation of the facility according to this proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated.
No. In general, the proposed amendment represents the conversion
of current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis
(Decommissioning Plan). Others represent minor curtailments of the
current requirements which are based on generic guidance or
previously approved provisions for other stations. These changes do
not involve revisions to the design of the station. Some of the
changes may involve revision in the operation of the station;
however, these provide additional restrictions which are in
accordance with the current safety analysis (Decommissioning Plan).
The proposed amendment for Dresden Station Unit 1's Technical
Specifications in general is based on STS guidelines or NRC accepted
changes to other facilities such as Trojan or San Onofre Unit 1. The
proposed amendment has been reviewed for acceptability at the
Dresden Nuclear Power Station considering similarity of system or
component design versus the STS of later operating plants. Any
deviations from STS requirements do not create the possibility of a
new of different kind of accident previously evaluated for Dresden
Station, Unit 1. No new modes of operation are introduced by the
proposed changes. The proposed changes maintain at least the present
level of operability. Therefore, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Will operation of the facility according to this proposed
change involve a significant reduction in a margin of safety?
No. In general, the proposed amendment represents the conversion
of current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis
(Decommissioning Plan). Others represent minor curtailments of the
current requirements which are based on generic guidance or
previously approved provisions for other stations. Some of the later
individual items may introduce minor reductions in the margin of
safety when compared to the current requirements. However, other
individual changes are the adoption of new requirements which will
provide significant enhancement of the reliability of human
performance assumed in the safety analysis (Decommissioning Plan),
or provide enhanced assurance that specified parameters remain
within their acceptance limits. These enhancements compensate for
the individual minor reductions, such that taken together, the
proposed changes will not significantly reduce the margin of safety.
The proposed amendment to Technical Specification Section 6.0
implements present requirements, or the intent of present
requirements in accordance with the guidelines set forth in the STS.
Any deviations from STS requirements do not significantly reduce the
margin of safety for Dresden Station. The proposed changes are
intended to improve readability, usability, and the understanding of
technical specification requirements while maintaining acceptable
levels of safe operation. The proposed changes have been evaluated
and found to be acceptable for use at Dresden based on system
design, safety analysis requirements and operational performance.
Since the proposed changes are based on NRC accepted provisions at
other operating plants that are applicable at Dresden and maintain
necessary levels of system or component reliability, the proposed
changes do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the analysis of the licensee 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: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450
Attorney for licensee: Michael I. Miller, Esquire, Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Seymour H. Weiss
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: December 2, 1996
Description of amendment request: The proposed amendments would
[[Page 4344]]
revise Technical Specification 3/4.4.2 to reduce the number of required
Safety/Relief Valves (SRVs). This change will support a modification to
remove five of the currently installed SRVs due to the current excess
capacity, and to reduce maintenance costs and worker radiation dose.
The current requirement for 17 of the 18 installed SRVs to be operable
would be changed to require 12 of the 13 installed SRVs to be operable.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The probability of an accident previously evaluated will not
increase as a result of this change, because the change in valve
configuration, and the accompanying piping modification does not
alter any of the initiators of an accident or cause them to occur
more frequently. The piping modifications will be performed
consistent with the current piping classifications for the affected
components. Removal of the SRVs will not impact the ability of the
remaining SRVs to perform their functions, as described below.
The consequences of an ASME Overpressurization Event are not
significantly increased and do not exceed the previously accepted
licensing criteria for this event. General Electric (GE) has
calculated the revised peak vessel pressure for LaSalle Station to
be 1341 psig, which is below the 1375 psig criterion of the ASME
Code for upset conditions, referenced in Section 5.2.2,
Overpressurization Protection, of the Updated Final Safety Analysis
Report (UFSAR), and NUREG-0519 (Safety Evaluation Report related to
the operation of LaSalle County Station, Units 1 and 2, March 1981),
and Section 15.2-4, Closure of Main Steam Isolation Valves (BWR) of
NUREG-0800 (Standard Review Plan). The consequences of this event
will continue to be verified on a cycle-specific basis, beginning
with LaSalle Unit 1 Cycle 9 (L1C9). These analysis results will be
approved as part of the normal reload licensing 10 CFR 50.59
processes.
GE has also performed an analysis of the limiting Anticipated
Transient Without Scram (ATWS) event, which is the MSIV Closure Event
(MSIVC). This analysis calculated the peak vessel pressure to be 1378
psig, which is well below the 1500 psig criterion of the ASME Code for
emergency conditions. General Electric has verified that these results
will not be impacted with the introduction of Siemens fuel.
The conclusions given in the safety analyses with regards to
primary containment dynamic loads, main steam piping loads, Loss-of-
Coolant Accident (LOCA) impact, Minimum Critical Power Ratio (MCPR)
impact and SRV availability also show that current accident and
transient analyses are not impacted by this change beyond those
reanalyzed by GE and discussed above.
There is no increase in the amount or types of radioactive
release for any of the affected accidents or transients.
Therefore, there is not a significant increase in the
consequences of an accident previously evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The as-left SRV piping configuration will continue to be
consistent with the current classifications for these piping and
supports, and have been evaluated by Sargent and Lundy analyses.
This ensures no different types of events may be caused by piping
failures at these locations. This is the only physical modification
proposed by this submittal, and it will not create the possibility
of a new or different kind of accident from those previously
evaluated. Other systems are not modified with this change and have
been shown in this submittal to continue to function as intended
with the new system configuration, with the exception of the
abandoned discharge line snubbers which may be replaced with struts,
except where they will be retained as snubbers due to thermal
expansion requirements. The changed supports are required to
function only as struts with the revised piping. Consideration and
evaluation of this function ensure no new or different accidents are
created.
3) Involve a significant reduction in the margin of safety
because:
While the calculated peak vessel pressures for the ASME
Overpressurization Event and the MSIVC ATWS Event are increased due
to the proposed SRV removals, the new peak pressures remain below
the respective licensing acceptance limits associated with these
events.
The actual cycle-specific reload analysis of the ASME
Overpressurization Event will be verified to be within the licensing
acceptance limit for that event prior to each cycle startup, as
required in the normal reload 10CFR50.59 process. These licensing
acceptance limits have been previously evaluated as providing a
sufficient margin of safety. For other accidents and transients,
including suppression pool loadings, the SRV removals have a
negligible, if any, effect on the results, so the margin of safety
is preserved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: August 14, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification Sections 3.3 (Engineered Safety
Features) and 6.9.1.9 (Core Operating Limits Report (COLR)); the basis
of Section 3.3, 3.6 (Containment) and 3.10 (Control Rods). These
changes would incorporate the best estimate approach into the licensing
basis for the Indian Point Unit No. 2 large break loss-of-coolant
accident (LOCA) analysis.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
No physical changes are being made by this change. The plant
conditions assumed in the analysis are bounded by the design
conditions for all equipment in the plant. Therefore, there will be
no increase in the probability of a loss-of-coolant accident. The
consequences of a LOCA are not being increased. That is, it is shown
that the emergency core cooling system is designed so that its
calculated cooling performance conforms to the criteria contained in
50.46 paragraph b, that is it meets the five criteria listed in
Section II [see application dated August 14, 1996] of this
evaluation. No other accident is potentially affected by this
change. Therefore, neither the probability nor the consequences of
an accident previously analyzed is increased due to the proposed
change.
2) Does the proposed license amendment create the possibility of
a new or different kind of accident from any previously analyzed?
Response:
There are no physical changes being made to the plant. No new
modes of plant operation are being introduced. The parameters
assumed in the analysis are within the design limits of existing
plant equipment. All plant systems will perform equally during the
response to a potential accident. Therefore, the possibility of a
new or different kind of accident than previously analyzed will not
be increased.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response:
It has been shown that the analytic technique used in the
analysis realistically describes the expected behavior of the Indian
Point Unit No. 2 reactor system during a
[[Page 4345]]
postulated loss of coolant accident. Uncertainties have been
accounted for as required by 10 CFR 50.46. A sufficient number of
loss of coolant accidents with different break sizes, different
locations and other variations in properties have been analyzed to
provide assurance that the most severe postulated loss of coolant
accidents were calculated. It has been shown by the analysis that
there is a high level of probability that all criteria contained in
10 CFR 50.46 paragraph b) are met. Therefore the proposed amendment
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Acting Director
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: August 21, 1996
Description of amendment request: The proposed amendment would
change the licensee's Technical Specifications (TSs) Section 3.3.G
(Hydrogen Recombiner System and Post-Accident Containment Venting
System), the basis for Section 3.3.G, and Section 4.4, Table 4.4-1
(Containment Isolation Valves). The change would remove the existing
flame-type hydrogen recombiners, its supporting equipment, and replace
it with passive autocatalytic recombiners (PARs). In addition, the
design basis analysis of post-accident hydrogen generation would be
recalculated.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Neither the probability nor the consequences of a post-LOCA
[loss-of-coolant accident] combustible gas accident are increased by
the change in recombiners or in the change to hydrogen generation
analysis. The probability of a 10 CFR 59.44 type LOCA is not
affected. The consequences of such an accident are not significantly
changed.
Accidents associated with failure of the flame-recombiner flue
(hydrogen/oxygen) system as well as with failure of the flame-
recombiner containment isolation valves have been eliminated.
No other accident is potentially affected by this change.
2) Does the proposed license amendment create the possibility of
a new or different kind of accident from any previously evaluated?
No new modes of plant operation are being introduced other than
elimination of operation of the flame-type recombiners and
associated support equipment. Recombiner failure is believed to be
far less likely with the PAR design but in any event, the
containment vent system is being maintained in its current role as
backup to recombiner systems. All other plant systems will perform
equally during the response to a potential accident. Therefore, the
possibility of a new or different kind of accident than previously
analyzed will not be increased.
3) Does the proposed amendment involve a significant reduction
in the margin of safety?
The proposed amendment involves margin in the hydrogen
flammability limit, in the hydrogen generation assumptions and in
the number of PAR devices assumed. Furthermore, sensitivity analysis
on PAR effectiveness indicates that additional margin exists for
success even with degraded PAR performance. It has been shown by the
analysis that the criteria of 10 CFR 50.44(d) can be met with
margin. Therefore, the proposed amendment does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Acting Director
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: August 22, 1996
Description of amendment request: The proposed amendment would
revise the licensee's Technical Specification Sections 3.3 and 4.5
(Engineered Safety Features). The proposed revision would delete the
requirement to utilize sodium hydroxide (NaOH) as an additive in the
posted-accident containment spray system.
Basis for proposed no significant hazards 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:
...consistent with the Commission's criteria in 10 CFR 50.92, we
have determined that the proposed change does not involve a
significant hazards consideration because the operation of Indian
Point Unit No. 2 in accordance with this change would not:
1) involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
revisions are based on conservative analyses utilizing new, approved
methodologies. The analysis shows the sodium hydroxide spray
additive can be removed without significantly affecting the
radiological consequences of a postulated LOCA [loss-of-coolant
accident] and that the calculated off-site doses would remain within
the 10 CFR 100 guidelines. In order to maintain acceptable pH levels
in the recirculating ECC [emergency core cooling] solution, baskets
of trisodium phosphate will be stored in strategic locations in
containment.
2) create the probability of a new or different kind of accident
from any accident previously evaluated. The proposed change allows
the containment safeguards to mitigate the consequences of a design
basis LOCA in a manner equivalent to that previously approved.
3) involve a significant reduction in a margin of safety. With
the proposed change, all safety criteria previously evaluated are
still met and remain conservative.
Therefore, based on the above, we conclude that the proposed
changes do not constitute a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Acting Director
Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of amendment request: January 3, 1997
Description of amendment request: The proposed amendments would
eliminate from various parts of the Technical Specifications any
requirement for the low steam pressure
[[Page 4346]]
signal as an initiator of safety injection. The licensee stated that
the function of the signal is adequately performed by other signals
(such as the low pressurizer pressure signal).
Basis for proposed no significant hazards 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 will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change, to delete the SI [safety injection] signal
on low steam line pressure, will only prevent an unnecessary SI
actuation as an event occurs which involves secondary system
depressurization. No consequences will significantly increase,
because for each event previously analyzed it has been shown that
either SI on low steam pressure is not demanded, or that another SI
signal (e.g., low pressurizer pressure) is generated in sufficient
time to meet applicable acceptance criteria. The probability of an
accident will not increase.
2. The proposed change will not create the possibility of any
new accident not previously evaluated.
The initiation of SI on a low steam line pressure signal may
occur during events which involve a depressurization of the
secondary side, including excessive auxiliary feedwater addition.
There are other SI initiation signals which will accomplish this
same function if needed. Removing this actuation signal will not
create any new failure modes or necessitate any new hardware
configurations (other than the deletion of the signal itself). No
new accident scenarios are created.
3. There is no significant reduction in a margin of safety.
Analysis has shown that for any transient for which SI would
have occurred on low steam line pressure, transient response is
maintained within acceptable limits. Steam line break mass and
energy releases inside containment do not violate the existing
environmental qualification envelope. Steam line breaks outside
containment are not adversely affected by this change.
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
proposed amendments involve 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: Herbert N. Berkow
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit Nos. 1 (ANO-1), Pope County, Arkansas
Date of amendment request: November 26, 1996
Description of amendment request: Change Reactor Coolant System
Pressure and Temperature Curvers
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Criterion 1 - Does not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change revises the pressure/temperature limits in
accordance with the 10 CFR 50.60 requirements or in accordance with
Code Case N-514. This approach utilizes the latest NRC guidelines
relative to estimating neutron irradiation damage of the reactor
vessel, as well as maintaining conservative limits with respect to
the low temperature overpressure protection (LTOP) system.
Therefore, this change does not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated since it
does not introduce new systems, failure modes or plant
perturbations. Therefore, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in Margin
of Safety.
The proposed change will not involve a significant reduction in
the margin of safety since the proposed pressure/temperature
limitations have been developed consistent with the requirements of
10 CFR 50.60. The operational limits have been developed to maintain
the necessary margins of safety through 32 effective full power
years using methodologies previously reviewed and approved by the
NRC. The objective of these limits is to prevent non-ductile failure
during any normal operating condition, including anticipated
operational occurrences and system hydrostatic tests.
The LTOP safety factors are based on reanalyzed conditions for
32 effective full power years of operation utilizing methodology
contained in ASME Code Case N-514. The LTOP evaluation under Code
Case N-514 for low temperature transients is considered more
appropriate than the ASME Section XI. The code case establishes a
factor of 110% of the pressure determined to satisfy Appendix G,
paragraph G-2215 of ASME Section XI, Division 1 as a design limit,
instead of 100% required by Section XI. This proposed alternative is
acceptable because the Code Case recognizes the conservatism of the
ASME Appendix G curves and allows establishing a LTOP setpoint which
retains an acceptable margin of safety while maintaining operational
margins for reactor coolant pump operation at low temperatures and
pressures. The Code Case provides an acceptable margin of safety
against flaw initiation and reactor vessel failure, and reduces the
potential for an undesired LTOP actuation. The application of Code
Case N-514 for ANO-1 will ensure an acceptable level of safety.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: October 7, 1996
Description of amendment request: Modify Plant Protection System
Test Interval to 123 days.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed changes included in this amendment request are
being made to surveillance intervals, allowances to use CISAM
elements and various administrative changes. These changes do not
alter the functional characteristics of any plant component and do
not allow any new modes of operation of any components. These
changes do not involve a significant increase in the probability of
any event initiator to occur. Therefore, this amendment request does
not involve a significant increase in the probability of any
accident previously evaluated.
Increasing the surveillance interval for the RPS and ESFAS
instrumentation has two principal effects with opposing impacts on
risk. The first impact is a slight increase in core damage frequency
that results from the increased unavailability of the
[[Page 4347]]
instrumentation in question from the extended testing interval. The
unavailability of the tested instrumentation components is
translated to result in a failure of the reactor to trip, an
anticipated transient without a scram, or a failure of the
appropriate engineered safety feature to actuate when required. The
opposing impact on risk is the corresponding reduction in core
damage frequency that would result due to the reduced exposure of
the plant to test induced transients.
Representative fault tree models were developed for ANO-2 and
the corresponding core damage frequency increases and decreases were
quantified in CEN-327 and CEN-327 Supplement 1. The NRC staff found
that changes in the RPS unavailabilities that result from extending
the surveillance test interval (STI) from 30 days to 90 days were
not considered to be significant. Estimates of the reduction in
scram frequency from the reduction in test induced scrams and the
corresponding reduction in core damage frequency were found
acceptable. Sequential testing intervals of 90 days were found to
result in a net reduction in risk.
CE NPSD-576 employed the same methodology used in CEN-327 and
its supplement to evaluate the impact of extending the surveillance
intervals from monthly sequential testing to every four months
(triannual) on a staggered test basis. The corresponding changes in
RPS and ESFAS unavailabilities are quantified in CE NPSD-576 and are
shown to be less than their counterparts in CEN-327 and its
supplement. Thus, triannual staggered testing should be acceptable
as it results in lower RPS and ESFAS unavailabilities than for a 90
day test interval with sequential testing which has been found to be
acceptable to the NRC.
The TS amendment request provided the option to use cycle
independent shape annealing matrix (CISAM) elements. The CISAM
elements will be validated during startup testing and will be
required to meet additional acceptance criteria as well as that used
for the cycle specific shape annealing matrix (SAM) elements. If the
CISAM is determined to be no longer valid, a cycle specific SAM will
be calculated and used in the CPCs. Therefore, the CPCs will operate
as designed and this change will not affect the consequences of any
accident previously evaluated.
The CPC addressable constant surveillance requirements and the
various administrative changes affected by this TS change do not
affect the consequences of any accident previously evaluated.
Therefore, these changes do not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
This amendment request does not involve any changes in equipment
and will not alter the manner in which the plant will be operated.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The RPS/ESFAS extended testing interval yields no significant
reduction in the margin to safety. The instrument drift occurring
over the proposed STI will not cause the setpoint values to exceed
those assumed in the safety analysis and specified in the TS. There
are no changes to equipment or plant operations that will result
from this change. The implementation of these proposed changes is
expected to result in an overall improvement in safety due to the
fact that reduced testing will result in fewer inadvertent trips,
less frequent actuation of EFAS components, and less frequent
distraction of the operations personnel.
The CPC addressable constant surveillance interval extension
included in this amendment request is consistent with the
methodology found in NUREG-1432, ``Standard Technical Specifications
Combustion Engineering Plants'' (ISTS). Requiring the addressable
constant verification to be performed as part of the CPC channel
functional test should detect an error in these constants prior to
restoring the channel to operable status instead of allowing the
error to go undetected until the next surveillance period. Although
the surveillance interval is extended by this TS change, this change
does not involve a significant reduction in the margin of safety.
The CPC CISAM elements and the various administrative changes
included in this TS change do not involve a significant reduction in
the margin of safety.
Therefore, these changes do not involve a significant reduction
in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: December 19, 1996
Description of amendment request: Change Request Concerning
Addition to the Core Operating Limit Report References
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Criterion 1 - Does not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change to add the technical manual for the
Combustion Engineering Nuclear Transient Simulation (CENTS) code to
the Core Operating Limits Report (COLR) references is administrative
in nature. The CENTS code has been reviewed and approved by the NRC.
The physical design or operation of the plant is not impacted by
this proposed change. The proposed change does not adversely impact
transient analysis assumptions or results. The COLR-related safety
analyses will continue to be performed utilizing NRC-approved
methodologies, and specific reload changes will be evaluated under
the provisions of 10CFR50.59. Therefore, this change does not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change to reference the NRC-approved CENTS code is
administrative in nature. No physical alterations of plant
configuration, changes to plant operating procedures, or operating
parameters are proposed. No new equipment is being introduced, and
no equipment is being operated in a manner inconsistent with its
design. The COLR-related safety analyses will continue to be
performed utilizing NRC-approved methodologies. A 10CFR50.59 safety
review will continue to be performed to evaluate specific reload
changes. Therefore, this change does not create the possibility of a
new or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed change to reference the CENTS code is
administrative in nature. Existing technical specification
operability and surveillance requirements are not reduced by the
proposed change. The cycle- specific COLR limits for future reloads
will continue to be developed based on NRC-approved methodologies.
Technical specifications will continue to require that the core be
operated within these limits and specify appropriate actions to be
taken if the limits are violated. The COLR-related safety analyses
will continue to be performed utilizing NRC-approved methodologies,
and specific reload changes will be evaluated per 10CFR50.59.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
[[Page 4348]]
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: December 19, 1996
Description of amendment request: Change Request Concerning Power
Calibration Requirements
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Criterion 1 - Does not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change will redefine the tolerance band allowed for
linear power level, the Core Protection Calculator (CPC) delta T
Power, and CPC nuclear power signals. Changing the tolerance range
from [plus or minus] 2% to between -0.5% and 10% between 15% and 80%
rated thermal power, will require more conservative tolerances than
are currently allowed. This change will ensure that the power
indications are more conservative relative to the existing safety
analyses. Therefore, this change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change to Technical Specification power calibration
tolerance limits are conservative relative to the current
requirements. This amendment request does not change the design or
operation of any plant systems or components. Therefore, this change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in Margin
of Safety.
The allowed tolerance band for the linear power level, CPC delta
T power, and CPC nuclear power signals between 15 and 80% power has
been redefined. The new requirements are more conservative than the
tolerances that currently exist in the Technical Specifications.
This change will ensure that the power indications are more
conservative relative to the existing safety analyses. Therefore,
this change does not involve a significant reduction in the margin
of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: December 19, 1996
Description of amendment request: Change Request Concerning Reactor
Coolant System Volume
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Criterion 1 - Does not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
This proposed change allows the relocation of the reactor
coolant system volume in the design features section of technical
specifications to the safety analysis report. Future changes will be
controlled under 10CFR50.59. This change is considered
administrative in nature. Appropriate values of reactor coolant
system volume are used in the safety analyses. This change does not
affect any system or component functional requirements. The
operation of the plant is not affected by this change.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The relocation of existing requirements from the technical
specifications to another licensee controlled document is
administrative in nature. This change does not modify or remove any
plant design requirement. The proposed change will not affect any
plant system or structure, nor will it affect any system functional
or operability requirements. Therefore, no new failure modes are
introduced as a result of this change.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed amendment request relocates the coolant system
volume located in the technical specifications design feature
section to another licensee controlled document, the ANO-2 Safety
Analysis Report, which is controlled under 10CFR50.59. The proposed
change is administrative in nature because the design requirements
for the facility remain the same. The proposed change does not
represent a change in the configuration or operation of the plant.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: December 19, 1996
Description of amendment request: Change Control Room Ventillation
System Requirements
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Criterion 1 - Does not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The control room emergency ventilation and air conditioning
systems are not initiators of an accident previously evaluated.
Extension of the allowable outage time for one inoperable control
room emergency air conditioning system from 7 days to 30 days is
acceptable based on the low probability of an event occurring that
would require control room isolation and a concurrent or subsequent
failure of the remaining operable control room emergency air
conditioning system. An evaluation using probabilistic safety
assessment techniques has shown the frequency of this event to be an
acceptably low level (4.67E-6/yr). The ANO-1 surveillance
requirements for the control room emergency ventilation and air
[[Page 4349]]
conditioning system have been updated for consistency with the ANO-2
requirements and are consistent with RG 1.52, March 1978, Revision 2
and ASTM D3803-1989. The change in the ANO-2 Mode of Applicability
for the control room radiation monitoring instrumentation is
acceptable because the only identified accident scenario requiring
control room isolation on high radiation while in Modes 5 and 6 is
the fuel handling accident and this analysis shows that the dose
consequences to the control room operators are acceptable in the
event of a fuel handling accident, assuming that the normal control
room ventilation system is properly isolated. The remainder of the
changes have been made for consistency between the ANO-1 and ANO-2
TS and are considered to be more restrictive or administrative in
nature.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The control room emergency ventilation and air conditioning
systems are not accident initiators. The proposed changes introduce
no new mode of plant operation and no new possibility for an
accident is introduced by modifying the ANO-1 surveillance testing
requirements for the control room emergency ventilation and air
conditioning systems.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
With the exception of the AOT extension and the relaxation of
the ANO-2 Mode of Applicability for the control room radiation
monitoring instrumentation, all the ANO-1 and ANO-2 changes are
considered administrative or more restrictive and are intended to
clarify and make consistent the requirements of the control room
emergency habitability equipment. Although the AOT extension does
involve an incremental reduction in the margin of safety due to
slight increase in the frequency of an event requiring control room
isolation, followed by failure of the operable emergency control
room chiller, a probabilistic safety assessment has shown this
slight increase in frequency (approximately 3.58E-6/yr) to be
acceptably low. The change in the ANO-2 Mode of Applicability for
the control room radiation monitoring instrumentation is acceptable
because the only identified accident scenario requiring control room
isolation on high radiation while in Modes 5 and 6 is the fuel
handling accident and this analysis shows that the dose consequences
to the control room operators are acceptable in the event of a fuel
handling accident, assuming that the normal control room ventilation
system is properly isolated.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket No.
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,
Georgia
Date of amendment request: December 3, 1996
Description of amendment request: The two proposed changes would
revise Technical Specification (TS) 2.1.1.2 for Hatch Nuclear Plant,
Unit 2, Safety Limit Minimum Critical Power Ratio (SLMCPR) values. The
revision is based upon unique plant evaluations for the current Cycle
13 and the use of General Electric (GE) GE-13 fuel, a 9 x 9 fuel
design, in the next Cycle 14. The proposed SLMCPRs for Hatch Unit 2 are
1.08 and 1.09 (single-loop operation) for the current Cycle 13, and
1.12 and 1.14 (single-loop operation) for Cycle 14.
The new SLMCPRs were calculated using NRC-approved methods and
interim implementing procedures. The SLMCPRs are set high enough to
ensure that greater than 99.9% of all fuel rods in the core avoid
transition boiling if the limit is not violated. The SLMCPRs
incorporate a margin for uncertainty in the core operating state for
uncertainties that are fuel-type dependent, including fuel bundle
nuclear characteristics, critical power correlation, and manufacturing
tolerances. These interim procedures were revised to incorporate the
following cycle-specific parameters: (1) Actual core loading, (2)
Conservative variations of projected control blade patterns, (3) Actual
bundle parameters (e.g., local peaking), and (4) Full cycle exposure
range.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration which is presented
below:
1. Does the change involve a significant increase in the
probability of consequences of an accident previously evaluated?
The derivation of the revised SLMCPRs for Plant Hatch Unit 2 for
incorporation into the Technical Specifications, and its use to
determine cycle-specific thermal limits, were performed using NRC-
approved methods. Additionally, interim implementing procedures
incorporating cycle-specific parameters were used. Based upon the
use of these calculations, revised SLMCPRs cannot increase the
probability or severity of an accident. The basis of the SLMCPR
calculation is to ensure that 99.9% of all fuel rods in
the core avoid transition boiling if the limit is not violated. The
new SLMCPRs preserve the existing margin to transition boiling and
fuel damage in the event of a postulated accident. Thus, it can be
concluded that the probability of fuel damage is not increased and
the proposed Technical Specifications changes do not involve an
increase in the probability or consequences of an accident
evaluation.
2. Do the proposed changes create the possibility of a new or
different type of accident from any previously evaluated?
The SLMCPR is a Technical Specifications numerical value
designed to ensure that fuel damage from transition boiling does not
occur as a result of the limiting postulated accident. The SLMCPRs
were calculated using NRC-approved methods. Additionally, interim
procedures incorporating cycle-specific parameters were used in the
analysis. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The margin of safety as defined in the Bases will remain the
same. The new SLMCPRs were calculated using NRC-approved methods
which are in accordance with the current fuel design and licensing
criteria. Additionally, interim implementing procedures, which
incorporate cycle-specific parameters were used. The SLMCPR remains
high enough to ensure that 99.9% of all fuel rods in the
core will avoid transition boiling if the limit is not violated,
thereby preserving the fuel cladding integrity. Therefore, the
proposed 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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
[[Page 4350]]
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of amendment request: January 7, 1997
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) for Plant Hatch, Units 1 and
2, associated with Surveillance Requirement (SR) testing that requires
manually actuating every safety/relief valve (S/RV) during each unit
startup from a refueling outage. The proposed changes would provide an
alternate method of testing the S/RVs during shutdown conditions rather
than during unit startup as is currently done. This approach would
reduce valve leakage, thereby reducing the possibility of inadvertent
valve actuation and resultant plant transients. Additionally, deletion
of testing for the safety mode of the S/RVs is proposed since other
testing provides operability verification.
Furthermore, the licensee proposes relief from the applicable
requirements of the ASME OM Code (1995), Appendix I, paragraph I
3.4.1(d), which also requires manual actuating of S/RVs during unit
startup.
Current Unit 1 and Unit 2 SRs 3.5.1.12 and 3.6.1.6.1 require that
each S/RV be manually actuated at pressure conditions. Georgia Power
Company (GPC) proposes to revise SRs 3.5.1.12 and 3.6.1.6.1 that would
require the S/RVs to be manually actuated in the relief mode during a
plant outage before steam is generated. The solenoid valve would be
energized, the actuator would stroke, and the pilot rod lift would be
measured. This in-situ test would verify that, given a signal to the
solenoid, the pilot disc rod would lift. If steam were present, the
pilot disc would open and initiate opening of the main stage.
The licensee also proposes to delete current Units 1 and 2 SR
3.4.3.2, which also requires that each S/RV be manually actuated
because this test is not necessary to assure S/RV operability in the
safety mode since other tests, taken together, confirm the entire S/RV
assembly functions adequately.
Basis for proposed no significant hazards 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:
Georgia Power Company [GPC] has reviewed the proposed license
amendment request and determined its adoption does not involve a
significant hazards consideration. In support of this determination,
an evaluation of each of the three 10 CFR 50.92 standards follows.
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Since the proposed Technical Specifications changes and ASME
Code relief do not impose any physical changes to the S/RVs, their
design function is unaffected. The submittal only proposes changes
to the manner in which the S/RVs are tested. As discussed in
Enclosure 1 [of the licensee's submittal], the combination of
current S/RV testing and the proposed alternate
testing will continue to adequately demonstrate the operability
of the S/RVs for both the safety and relief modes. Under the
proposed testing requirements, it is expected that S/RV leakage will
decrease; thus, the probability of occurrence of an inadvertent S/RV
actuation is actually reduced.
FSAR [Final Safety Analysis Report] analyzed events, such as
MSIV [main steam isolation valve] closure, generator load reject,
turbine trip with failure of switchyard breakers to open, and
pressure regulator failure, take credit for the S/RVs mitigating the
consequences of these events. These proposed changes will not
increase the consequences of these events, since a series of S/RV
tests (on the bench and installed) will ensure all S/RV components
necessary to ensure valve opening will function. The S/RVs will
therefore be capable of performing their design functions.
Furthermore, reducing the number of manual actuations of the S/
RVs decreases the likelihood of a stuck open S/RV, which is an
analyzed event in the Hatch FSAR.
Therefore, the probability of occurrence and the consequences of
previously analyzed events are not increased.
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 affect the manner in which S/RV operability
is verified in that one Technical Specifications SR [surveillance
requirement] is being deleted and two are being revised; however,
they do not affect the way the S/RVs are operated. The S/RVs will
not be operated or tested in a manner contrary to their design. As a
result, no new mode of operation is introduced. That is, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The present method of S/RV testing unnecessarily challenges the
valves, and is linked to S/RV degradation through pilot valve and/or
main valve leakage. This Technical Specifications change should
decrease S/RV leakage and improve S/RV reliability by reducing the
potential for spurious valve actuation at full power. In this sense,
the margin of safety is actually increased; e.g., the likelihood for
spurious S/RV actuation is reduced.
Deleting the test of installed S/RVs at rated temperature and
pressure will not significantly reduce the margin of safety for
events in which S/RV actuation is assumed, since each S/RV receives
a series of tests which insure each component necessary for
successful opening of the S/RV functions properly. Thus, the S/RV is
assured of opening in either the safety or the relief mode. For
example, at Wyle Labs, the valves undergo testing at operating steam
pressure. This test ensures operability of the pilot and main discs
and also verifies set pressure, reseat pressure, and main steam
stroke time. As noted previously, upon successful completion of
these tests, including verification of zero seat leakage, the valves
receive a written certification from the lab and are returned to
Plant Hatch for installation.
GPC further proposes that, upon installation, but before steam
is generated, the valves receive a test requiring the solenoid to be
energized. This test provides additional verification that the pilot
disc opens. The remaining segments of the S/RV tests verify the
ability of ADS and LLS logic to energize the solenoid.
In summary, this amendment does not involve a significant
reduction in the margin of safety, because of the reduction in S/RV
degradation, and because remaining tests confirm the valves will
function properly when required.
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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
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 16, 1996
Description of amendment request: The amendment request, if
approved, would reflect the change in the legal name of the operator of
TMI-1 from GPU Nuclear Corporation to GPU Nuclear Inc. and reflect in
the TMI-1 license and the Technical Specifications the registered trade
name of GPU Energy.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration (SHC), which is
presented below:
GPU Nuclear Inc. has determined that the proposed TMI-1 license
amendment and technical specification change request
[[Page 4351]]
involve no significant hazards consideration as defined in 10 CFR
50.92 because:
Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment adds to the license and
the technical specifications the trade name of the Owners of TMI-1.
The change in the legal name of the operator of TMI-1 is a cosmetic
change made to reflect the name changes made throughout the GPU
family of companies. The name change has no impact on plant design
or operation.
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 because no
new failure modes are created by the proposed changes. The use of a
common trade name for the Owners of TMI-1 and the change in the
legal name of the operator of TMI-1 has no impact on plant design or
operation. Thus, there is no creation of the possibility of a new or
different kind of accident from those previously evaluated.
Operation of the facility in accordance with the proposed
amendment will not involve a signficant reduction in a margin of
safety. The proposed amendment does not change any operating limits
for reactor operation.
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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Acting Project Director: Patrick D. Milano
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit No. 1, Berrien County, Michigan
Date of amendment request: August 4, 1995 as supplemented December
20, 1996 [AEP:NRC:1129E and 1129M]
Description of amendment request: The proposed amendment would
modify the technical specifications (T/S) to allow for repair of hybrid
expansion joint (HEJ) sleeves under redefined repair boundary limits.
This alternate plugging criterion would assess the integrity of parent
tube indications based on the degraded joint geometry, with reference
to the specific location of the flaw. The continued operability of the
HEJ sleeved tube would be based on the measured diameter difference, or
diameter delta (delta D), between the sleeve peak hardroll diameter and
the diameter of the sleeve adjacent to the parent tube flaw in the
upper joint.
Basis for proposed no significant hazards 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:
Conformance of the proposed amendments to the standards for a
determination of no significant hazard as defined in 10 CFR 50.92
(three factor test) is shown in the following.
(1) Operation of 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.
The HEJ sleeved tube structural integrity limits defined by this
amendment provide for structural integrity consistent with the
guidance of RG 1.121. Tube structural integrity consistent with the
most limiting RG 1.121 loading is inherently provided by a measured
[delta D] of less than 1 mil, although the criterion specifies a
minimum of 3 mils must be verified. The structural integrity
characteristics of a postulated degraded parent tube with a 3 mil
[delta D] provides for axial restraint capability of more than
double the most limiting RG 1.121 loading, which indicates that the
postulated separated tube would not become axially displaced
relative to the sleeve during any plant condition.
Based on tube pull data from Cook Nuclear Plant and other plants
it is expected that TSP intersections would provide a substantial
axial restraint capability. This interaction is neglected in the
analysis of the criterion, and provides for extra safety margin.
Based on the destructive examination results for sections of HEJ
sleeved tubes removed in 1994 from another plant, the parent tube
flaw morphology is described as circumferentially oriented with
multiple initiation sites. This segmented morphology indicates that
the previously performed structural capability testing is
conservative. Additional axial load bearing capability is provided
by the segmented morphology since end cap loading would be
transmitted through the tube by the non-degraded ligaments of the
segmented crack network, and tube separation therefore, is not
likely or credible.
The consequences of any postulated failure of a sleeved tube to
which the criteria has been applied would be bounded by the current
steam generator tube rupture event discussed in the Cook Nuclear
Plant Final Safety Analysis Report (FSAR). Axial displacement of any
tube, sleeved or unsleeved, is bounded by approximately 1.1 inch. A
tube which experiences axial displacement by this amount would be
expected to exhibit a release rate well below the normal makeup
capacity. In order for a HEJ sleeved tube to exhibit reactor coolant
system release rates approaching the release rates assumed in the
FSAR the tube must be displaced by approximately 3 inches. In order
for the postulated separated tube to experience axial displacement
of any magnitude, it must be assumed that the HEJ hardroll provides
no structural benefit and that the tube-to-TSP interaction is
frictionless.
Postulated primary to secondary leakage during a main steam line
break event will be assessed against the limit of 8.4 gpm in the
faulted loop, calculated as part of the voltage based plugging limit
for tube support plate intersections. The total of all leakage
sources must be shown to be less than this value.
Application of the 3 mil [delta D] criterion (excluding eddy
current uncertainty) does not change existing reactor coolant system
flow conditions, therefore, existing LOCA analysis results will be
unaffected. Plant response to design basis accidents for the current
tube plugging and flow conditions are not affected by the repair
process; no new tube diameter restriction is introduced.
(2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Application of the proposed 3 mil [delta D] HEJ sleeved tube
structural integrity criterion will not introduce significant or
adverse changes to the plant design basis. The 3 mil [delta D]
criteria provides for structural integrity of the HEJ sleeved tube
assembly which significantly exceeds the limiting RG 1.121 loading
condition. Under these conditions neither a single nor a multiple
tube rupture event is considered credible.
The general outline of the HEJ sleeve is unaffected, and the
application of the proposed criterion does not change the sleeve
configuration or size/shape. The application of the criterion also
does not represent a potential to affect other plant components.
(3) The proposed license amendment does not involve a
significant reduction in a margin of safety.
The proposed criterion has been shown to provide structural
integrity of the tube bundle consistent with the most limiting RG
1.121 tube integrity recommendations. In order for tube rupture to
occur, the degraded parent tube must experience a complete
circumferential separation and be subsequently axially displaced by
approximately 3 inches. The inherent structural integrity provided
by the interference fit of the HEJ in addition to the axial
restraint provided by tube support plate intersections above the HEJ
provides for structural integrity far exceeding the RG 1.121 loading
of 2264 lb. Even in the event that a degraded HEJ sleeved parent
tube were to experience axial displacement, the maximum amount of
displacement the tube could experience is bounded by 1.11 inch.
Postulating that the tube were to become displaced by this amount,
primary to secondary leakage would be limited to well less than the
normal makeup capacity due to the proximity between the
hydraulically expanded sleeve OD and tube ID.
Pulled HEJ sleeved tube samples from another plant with HEJ
sleeved tubes indicate that the crack morphology is described as
circumferentially oriented cracking with multiple initiation sites.
This segmented
[[Page 4352]]
morphology provides for additional structural margin not modeled in
the testing program.
Existing flow equivalency calculations for the HEJ sleeved tubes
will be unaffected by the application of the criterion.
Based on the preceding analysis it is concluded that operation
of Cook Nuclear Plant unit 1 following the application of the 3 mil
[delta D] HEJ sleeved tube structural integrity limit does not
increase the probability of an accident previously evaluated, create
the possibility of a new or different kind of accident from any
accident previously evaluated, or reduce any margins to plant
safety. Therefore, the license amendment does not involve a
significant hazards consideration as defined in 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: 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: Gail H. Marcus
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit
No. 3, York County, Pennsylvania
Date of application for amendment: October 30, 1996
Description of amendment request: These amendments revise the
safety limit minimum critical power ratios (SLMCPRs) at Peach Bottom
Atomic Power Station, Unit 3.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1) The proposed TS [technical specification] changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The derivation of the cycle-specific SLMCPRs for incorporation
into the TS, and its use to determine cycle-specific thermal limits,
have been performed using USNRC [U.S. Nuclear Regulatory
Commission]-approved methods as discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-11, and U.S.
Supplement, NEDE-24011-P-A-11-US, November 17, 1995 and interim
(reconfirmation) implementing procedures. This change in SLMCPRs
cannot increase the probability or severity of an accident.
The basis of the SLMCPR calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid transition boiling if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling and fuel damage in the event of a
postulated accident. The fuel licensing acceptance criteria for the
SLMCPR calculation apply to PBAPS [Peach Bottom Atomic Power
Station], Unit 3, Cycle 11 in the same manner as they have applied
previously. The probability of fuel damage is not increased.
Therefore, the proposed TS changes do not involve an increase in the
probability or consequences of an accident previously evaluated.
2) The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SLMCPR is a TS numerical value, designed to ensure that
transition boiling does not occur in 99.9% of all fuel rods in the
core during the limiting postulated accident. It cannot create the
possibility of any new type of accident. The new SLMCPRs are
calculated using USNRC-approved methods (General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-11, and U.S.
Supplement, NEDE-24011-P-A-11-US, November 17, 1995) and interim
(reconfirmation) implementing procedures.
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 margin of safety as defined in the TS Bases will remain the
same. The new SLMCPRs are calculated using USNRC-approved methods
(General Electric Standard Application for Reactor
Fuel,'' NEDE-24011-P-A-11, and U.S. Supplement, NEDE-24011-P-A-11-
US, November 17, 1995) and interim (reconfirmation) implementing
procedures which are in accordance with the current fuel licensing
criteria. The SLMCPRs ensure that greater than 99.9% of all fuel
rods in the core will avoid transition boiling if the limit is not
violated, thereby preserving the fuel cladding integrity. Therefore,
the proposed TS 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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101
NRC Project Director: John F. Stolz
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: January 15, 1997
Description of amendment request: The amendment proposes to
relocate the snubber operability, surveillance, and record requirements
for components (snubbers) in the Technical Specifications (TS) to plant
controlled documents.
Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, based on the following:
1. These changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated
because:
The changes relocate operability, surveillance, and record
requirements for components (snubbers) which do not meet the
criteria for inclusion in the Technical Specifications (TS). The
affected components are not assumed to be initiators of analyzed
events and are not assumed to mitigate accident or transient events.
The snubber requirements will be relocated from the TS to plant
controlled documents. These requirements will be maintained pursuant
to 10 CFR 50.59. Therefore, the changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The changes do not create the possibility of a new or
different type of accident previously evaluated because:
The changes do not necessitate a physical alteration of the
plant (no new or different type of equipment will be installed) or
affect parameters governing normal plant operation. Adequate control
of future changes to snubber requirements will be maintained. Thus,
these changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated for the
plant.
3. The proposed changes do not involve a reduction in a margin
of safety because:
The changes do not involve a change to the operability,
surveillance, and record requirements for the snubber program as
they currently exist in the TS, nor do they impact on any safety
analysis assumptions. The proposed changes relocate snubber
requirements from the TS to plant controlled documents. Changes to
the requirements in these documents are subject to the requirements
of 10 CFR 50.59. In addition, exceptions to code requirements for
testing will require NRC approval. Regulations and
[[Page 4353]]
FitzPatrick commitments to the NRC contain the necessary
programmatic requirements for the plant controlled documents.
Operating limitations will continue to be imposed, and required
surveillances will continue to be performed in accordance with
regulations, FitzPatrick commitments to the NRC, and written
procedures and instructions that are auditable by the NRC. If
snubber inoperability causes a TS system or component to be
inoperable, then the affected system or component Limiting Condition
for Operation (LCO) will be entered. Based on the above, 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Acting Director
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: January 7, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.2.5 to incorporate an exception
to the provisions of TS 4.0.4 and to clarify the time at which the
surveillance can be performed by adding that the surveillance is to be
performed within 24 hours after attaining steady state conditions at or
above 90% rated thermal power. The revised surveillance would also
contain editorial enhancements that do not change the intent of the
current surveillance. TS Table 3.2-1 for Salem Unit 1 would be revised
to delete reference to three loop operation (which is not permitted at
Salem Unit 1) in order to eliminate potential confusion when applying
this table.
Basis for proposed no significant hazards 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 changes proposed on the RCS [Reactor Coolant System] flow
measurement and exemption to Specification 4.0.4 do not affect the
operation of the equipment during conditions when they are required
to perform their safety function. No physical changes to the plant
result from the proposed changes made to the surveillance
requirements. The measurement of RCS flow does not impact the
probability of an accident.
Testing is being performed with the plant in the condition in
which the automatic initiation signals for low RCS flow would result
in a time consistent with the TS requirements.
Protection System in providing a reactor trip upon a loss of RCS
flow. Degradations in flow will occur over a long duration; however,
testing will continue to be performed within twenty-four hours upon
achieving steady state greater than or equal to 90% RTP [Rated
Thermal Power] after refueling which is a sufficiently short
duration after startup to identify flow degradations.
Changes proposed to refer to Table 3.2-1 for the DNB [Departure
from Nucleate Boiling] parameters and to delete the Unit 1 three
loop operation parameters, and the inclusion of the type of test
performed are editorial in nature.
Therefore, the consequences of an accident previously evaluated
are not significantly increased by the proposed changes.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve any modifications to
existing plant equipment, do not alter the function of any plant
systems, do not introduce any new operating configurations or new
modes of plant operation, nor change the safety analyses. The point
at which RCS flow is measured using a heat balance will not impact
the ability to maintain or monitor Reactor Coolant flows. The
proposed changes will, therefore, not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Changes proposed to refer to Table 3.2-1 for the DNB parameters
and to delete the Unit 1 three loop operation parameters, and the
inclusion of the type of test performed are editorial in nature.
[The proposed changes will, therefore, not create the
possibility of a new or different kind of accident from any accident
previously evaluated.]
3. The proposed change does not involve a significant reduction
in a margin of safety.
The changes to the RCS flow surveillance do not decrease the
scope of the existing testing, but will clarify the point at which
the testing is performed.
The time in which testing is performed, after achieving steady
state conditions after reaching greater than or equal to 90% RTP
ensures that testing is performed in a timely manner. Flow margins
established as a result of previous testing will not be
significantly reduced in light of recent outage activities. Future
changes that might impact margins established by the testing will be
reviewed in accordance with the requirements of 10 CFR 50.59.
Changes proposed to refer to Table 3.2-1 for the DNB parameters
and to delete the Unit 1 three loop operation parameters, and the
inclusion of the type of test performed are editorial in nature.
All changes are consistent with the intent of Salem's current TS
[Technical Specification] and with the 18 month surveillances
specified in NUREG-1431, Revision 1.
The proposed change, therefore, 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: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendments request: December 26, 1996
Description of amendments request: The proposed amendment would
revise Technical Specification 3/4.4.6 ``Steam Generators'' and its
associated Bases. Specifically, the steam generator repair limit would
be modified to clarify that the appropriate method for determining
serviceability for tubes with outside diameter stress corrosion
cracking at the tube support plate is by a methodology that more
reliably assesses structural integrity. This amendment request is in
accordance with NRC's Generic Letter 95-05, ``Voltage-Based Repair
Criteria for Westinghouse Steam Generator Tubes Affected by Outside
Diameter Stress Corrosion Cracking.''
Basis for proposed no significant hazards 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 Farley units 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 standing tubes at
room temperature conditions shows burst pressures as high as
approximately 5000 psi for indications of outer diameter stress
corrosion cracking with voltage measurements as high as 26.5 volts.
Burst testing performed on pulled tubes,
[[Page 4354]]
including tubes pulled from Farley Unit 1, with up to 7.5 volt
indications show burst pressures in excess of 5300 psi at room
temperature. ... [T]ube burst criteria are inherently satisfied
during normal operating conditions by the presence of the tube
support plate. Furthermore, 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 R.G. [Regulatory Guide] 1.121
criterion requiring the maintenance of a margin of 1.43 times the
steam line break pressure differential on tube burst if through-wall
cracks are present without regard to the presence of the tube
support plate. Considering the existing data base, this criterion is
satisfied with bobbin coil indications with signal amplitudes over
twice the 2.0 volt voltage-based repair criteria, regardless of the
indicated depth measurement. This structural limit is based on a
lower 95% confidence level limit of the data at operating
temperatures. The 2.0 volt criterion provides an extremely
conservative margin of safety to the structural limit considering
expected growth rates of outside diameter stress corrosion cracking
at Farley. Alternate crack morphologies can correspond to a voltage
so that a unique crack length is not defined by a burst pressure to
voltage correlation. However, relative to expected leakage during
normal operating conditions, no field leakage has been reported from
tubes with indications with a voltage level of under 7.7 volts for a
3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as
compared to the 2.0 volt proposed voltage-based tube repair limit).
Thus, the proposed amendment does not involve a significant increase
in the probability or consequences of an accident.
Relative to the expected leakage during accident condition
loadings, the accidents that are affected by primary-to-secondary
leakage and steam release to the environment are Loss of External
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator
Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a
Control Rod Drive Mechanism Housing. Of these, the Major Secondary
System Pipe Failure is the most limiting for Farley in considering
the potential for off-site doses. The offsite dose analyses for the
other events which model primary-to-secondary leakage and steam
releases from the secondary side to the environment assume that the
secondary side remains intact. The steam generator tubes are not
subjected to a sustained increase in differential pressure, as is
the case following a steam line break event. This increase in
differential pressure is responsible for the postulated increase in
leakage and associated offsite doses following a steam line break
event. In addition, the steam line break event results in a bypass
of containment for steam generator leakage. Upon implementation of
the voltage-based repair criteria, it must be verified that the
expected distributions of cracking indications at the tube support
plate intersections are such that primary-to-secondary leakage would
result in site boundary dose within the current licensing basis.
Data indicate that a threshold voltage of 2.8 volts could result in
through-wall cracks long enough to leak at steam line break
conditions. Application of the proposed repair criteria requires
that the current distribution of a number of indications versus
voltage be obtained during the refueling outages. The current
voltage is then combined with the rate of change in voltage
measurement and a voltage measurement uncertainty to establish an
end of cycle voltage distribution and, thus, leak rate during steam
line break pressure differential. The leak rate during a steam line
break is further increased by a factor related to the probability of
detection of the flaws. If it is found that the potential steam line
break leakage for degraded intersections planned to be left in
service coupled with the reduced allowable specific activity levels
result in radiological consequences outside the current licensing
basis, then additional tubes will be plugged or repaired to reduce
steam line break leakage potential to within the acceptance limit.
Thus, the consequences of the most limiting design basis accident
are constrained to present licensing basis limits, and therefore
there is no change to the probability or consequences of an accident
previously evaluated.
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 voltage-based tube repair
criteria does not introduce any significant changes to the plant
design basis. Use of the criteria does not provide a mechanism that
could result in an accident outside of the region of the tube
support plate elevations. Neither a single or multiple tube rupture
event would be expected in a steam generator in which the repair
criteria have been applied during all plant conditions. The bobbin
probe signal amplitude repair criteria are established such that
operational leakage or excessive leakage during a postulated steam
line break condition is not anticipated. Southern Nuclear has
previously implemented a maximum leakage limit of 140 gpd per steam
generator. The R.G. 1.121 criterion for establishing operational
leakage limits that require plant shutdown are based upon leak-
before-break considerations to detect a free span crack before
potential tube rupture. The 140 gpd limit provides 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. R.G. 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 factor
of safety of 1.43 against bursting at steam line break pressure
differential. A voltage amplitude of approximately 9 volts for
typical outside diameter stress corrosion cracking corresponds to
meeting this tube burst requirement at the 95% prediction interval
on the burst correlation. Alternate crack morphologies can
correspond to a voltage so that a unique crack length is not defined
by the burst pressure versus voltage correlation. Consequently, a
typical burst pressure versus through-wall crack length correlation
is used below to define the ``longest permissible crack'' for
evaluating operating leakage limits.
The single through-wall crack lengths that result in tube burst
at 1.43 times steam line break pressure differential and steam line
break conditions are about 0.54 inch and 0.84 inch, respectively.
Normal leakage for these crack lengths would range from about 0.4
gallons per minute to 4.5 gallons per minute, respectively, while
lower 95% confidence level leak rates would range from about 0.06
gallons per minute to 0.6 gallons per minute, respectively.
An operating leak rate of 140 gpd per steam generator has been
implemented. This leakage limit provides for detection of 0.4 inch
long cracks at nominal leak rates and 0.6 inch long cracks at the
lower 95% confidence level leak rates. Thus, the 140 gpd limit
provides for plant shutdown prior to reaching critical crack lengths
for steam line break conditions at leak rates less than a lower 95%
confidence level and for three times normal operating pressure
differential at less than nominal leak rates.
Considering the above, the implementation of voltage-based
repair criteria will not create the possibility of a new or
different kind of accident from any previously evaluated.
3) The proposed license amendment does not involve a significant
reduction in margin of safety.
The use of the voltage-based repair criteria is demonstrated to
maintain steam generator tube integrity commensurate with the
requirements of Generic Letter 95-05 and R.G. 1.121. R.G. 1.121
describes a method acceptable to the NRC staff for meeting GDC
[General Design Criteria] 2, 14, 15, 31, and 32 by reducing the
probability of 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 outside diameter
stress corrosion cracking at the tube support plate elevations is
not expected to lead to a steam generator tube rupture event during
normal or faulted plant conditions. The most limiting effect would
be a possible increase in leakage during a steam line break event.
Excessive leakage during a steam line break event, however, is
precluded by verifying that, once the criteria are applied, the
expected end of cycle distribution of crack indications at the tube
support plate elevations would result in minimal, and acceptable
primary to secondary leakage during the event and, hence, help to
demonstrate radiological conditions are less than an appropriate
fraction of the 10 CFR [Part] 100 guideline.
The margin to burst for the tubes using the voltage-based repair
criteria is comparable to that currently provided by existing
technical specifications.
In addressing the combined effects of LOCA [loss-of-coolant
accident] + SSE [safe shutdown earthquake] on the steam generator
[[Page 4355]]
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 tube support plates may become deformed as a result
of lateral loads at the wedge supports at the periphery of the plate
due to either the LOCA rarefaction wave and/or 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 [reactor coolant system] flow area through the tubes. The
reduction in flow area 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 the
partial through-wall cracks in tubes could progress to through-wall
cracks during tube deformation or collapse or that short through-
wall indications would leak at significantly higher leak rates than
included in the leak rate assessments.
Consequently, a detailed leak-before-break analysis was
performed and it was concluded that the leak-before-break
methodology (as permitted by GDC 4) is applicable to the Farley
reactor coolant system primary loops and, thus, the probability of
breaks in the primary loop piping is sufficiently low that they need
not be considered in the structural design basis of the plant.
Excluding breaks in the RCS primary loops, the LOCA loads from the
large branch line breaks were analyzed at Farley and were found to
be of insufficient magnitude to result in steam generator tube
collapse or significant deformation.
Regardless of whether or not leak-before-break is applied to the
primary loop piping at Farley, any flow area reduction is expected
to be minimal (much less than 1%) and PCT margin is available to
account for this potential effect. Based on analyses' results, no
tubes near wedge locations are expected to collapse or deform to the
degree that secondary to primary in-leakage would be increased over
current expected levels. For all other steam generator tubes, the
possibility of secondary-to-primary leakage in the event of a LOCA +
SSE event is not significant. In actuality, the amount of secondary-
to-primary leakage in the event of a LOCA + SSE is expected to be
less than that originally allowed, i.e., 500 gpd per steam
generator. Furthermore, secondary-to-primary in-leakage would be
less than primary-to-secondary leakage for the same pressure
differential since the cracks would tend to tighten under a
secondary-to-primary pressure differential. Also, the presence of
the tube support plate is expected to reduce the amount of in-
leakage.
Addressing the R.G. 1.83 considerations, implementation of the
tube repair criteria is supplemented by 100% inspection requirements
at the tube support plate elevations having outside diameter stress
corrosion cracking indications, reduced operating leakage limits,
eddy current inspection guidelines to provide consistency in voltage
normalization, and rotating probe inspection requirements for the
larger indications left in service to characterize the principle
degradation mechanism as outside diameter stress corrosion cracking.
As noted previously, implementation of the voltage-based repair
criteria will decrease the number of tubes that must be taken out of
service with tube plugs or repaired. The installation of steam
generator tube plugs or tube sleeves would reduce the RCS flow
margin, thus implementation of the voltage-based repair criteria
will maintain the margin of flow that would otherwise be reduced
through increased tube plugging or sleeving.
Considering the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the Final Safety Analysis Report or any
bases 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendments request: January 10, 1997
Description of amendments request: The proposed amendments would
implement repair of tubes using laser welded tube sleeves for the steam
generators at Farley Units 1 and 2 as described in WCAP-13088, Revision
4, and WCAP-14740. In addition, for Unit 2, references to a one-cycle
limited implementation of L* are being removed. The approval for the
limited implementation of L* expired at the last Unit 2 outage in the
fall of 1996.
Basis for proposed no significant hazards 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 Farley Units 1 and 2 in accordance with the
proposed license amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The laser welded sleeve configurations as described within WCAP-
13088, Revision 4 and WCAP-14740 have been designed and analyzed in
accordance with the requirements of the ASME Code [American Society
of Mechanical Engineers Boiler and Pressure Vessel Code]. Fatigue
and stress analyses of the sleeved tube assemblies produced
acceptable results. Mechanical testing has shown that the structural
strength of Alloy 690 sleeves under normal, faulted and upset
conditions is within acceptable limits. Leakage testing for 7/8 inch
tube sleeves has demonstrated that significant primary-to-secondary
leakage is notexpected during all plant conditions, including the
case where the seal weld is not produced in the lower joint of the
tubesheet sleeve.
Initial acceptance of welded joints uses ultrasonic inspection
to verify that all weld thicknesses meet the minimum specified
conditions over the entire circumference. A plugging limit of 24%
allowable depth of penetration of the sleeve tube wall thickness
applies for each type of laser welded sleeve that may be installed
in the Farley Nuclear Plant steam generators and is determined for
uprated conditions with a limiting steam pressure for reduced
Thot and 20% steam generator tube plugging conditions. These
conditions represent the limiting primary-to-secondary operating
pressure differential, which is bounding for the sleeve plugging
limit and structural analysis inputs. However, the state-of-the-art
in eddy current inspection capability is such that no probes are
qualified to size the depth of penetration of stress corrosion
cracking. It is generally believed that the detection threshold of
these probes is well below 40% throughwall. Southern Nuclear
Operating Company will plug on detection any crack-like indications
that may occur in the sleeve using the sleeve inspection probe of
record until an inspection process is qualified to size depth of
penetration of stress corrosion cracking into the tube wall.
The hypothetical consequences of failure of the sleeve would be
bounded by the current steam generator tube rupture analysis
included in the Farley Nuclear Plant FSAR [Final Safety Analysis
Report]. Due to the slight reduction in diameter caused by the
sleeve wall thickness, it is expected that primary coolant release
rates would be slightly less than assumed for the steam generator
tube rupture analysis (depending on the break location), and
therefore, would result in lower total primary fluid mass release to
the secondary system. Combinations of tubesheet sleeves and tube
support plate sleeves would reduce the primary fluid flow through
the sleeved tube assembly due to the series of diameter reductions
the fluid would have to pass on its way to the break area. The
overall effect would be reduced steam generator tube rupture release
rates.
As addressed previously, the proposed Technical Specification
change to support the installation of full length tubesheet,
elevated tubesheet, or tube support plate elevation Alloy 690 laser
welded sleeves as described in WCAP-13088, Revision 4 and WCAP-14740
does not adversely impact any other previously evaluated design
basis accident or the results of LOCA [loss-
of-coolant accident] and non-LOCA accident analyses for the
current Technical Specification minimum reactor coolant
[[Page 4356]]
system flow rate. The results of the analyses and testing, as well
as plant operating experience, demonstrate that the sleeve assembly
is an acceptable means of restoring tube integrity to a condition
consistent with its original design basis. Also, per Regulatory
Guide 1.83, Revision 1 recommendations, the condition of the sleeved
tube can be monitored through periodic inspections with present eddy
current techniques.
Conformance of the sleeve design with the applicable sections of
the ASME Code and results of the leakage and mechanical tests
support the conclusion that the installation of laser welded tube
sleeves will not increase the probability or consequences of an
accident previously evaluated. Depending upon the break location for
a postulated steam generator tube rupture event, implementation of
tube sleeving could act to reduce the radiological consequences to
the public due to reduced primary to secondary flow rate through a
sleeved tube compared to a non-sleeved tube based on the restriction
afforded by the sleeve wall thickness.
Removal of the references to the interim use of an L* repair
criteria will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
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 laser welded sleeving will not introduce
significant or adverse changes to the plant design basis. Sleeving
also does not represent a potential to affect any other plant
component. Stress and fatigue analysis of the repair has shown the
ASME Code minimum stress values are not exceeded. Implementation of
laser welded sleeving maintains overall tube bundle structural and
leakage integrity at a level consistent to that of the originally
supplied tubing during all plant conditions. Leak and mechanical
testing of sleeves support the conclusions of the calculations that
each sleeve joint retains both structural and leakage integrity
during all conditions. Sleeving of tubes does not provide a
mechanism resulting in an accident outside of the area affected by
the sleeves. Any hypothetical accident as a result of potential tube
or sleeve degradation in the repaired portion of the tube is bounded
by the existing tube rupture accident analysis. Since the sleeve
design does not affect any other component or location of the tube
outside of the immediate area repaired, in addition to the fact that
the installation of sleeves and the impact on current plugging level
analyses is accounted for, the possibility that laser welded
sleeving creates a new or different type of accident is not
credible.
Removal of the references to the interim use of an L* repair
criteria will 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 margin of safety.
The laser welded sleeving repair of degraded steam generator
tubes as identified in WCAP-13088, Revision 4, has been shown by
analysis to restore the integrity of the tube bundle consistent with
its original design basis condition as the requirements of the ASME
Code are satisfied. The safety factors used in the design of sleeves
for the repair of degraded tubes are consistent with the safety
factors in the ASME Boiler and Pressure Vessel Code used in steam
generator design. The design of the tubesheet sleeve lower joints
for the 7/8 inch sleeves (for both the full length and elevated
tubesheet sleeve) have been verified by testing to preclude
realistic leakage during normal and postulated accident conditions.
The portions of the installed sleeve assembly which represent
the reactor coolant pressure boundary can be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the recommendations of Regulatory Guide 1.83, Revision 1
and the surveillance requirements included in Specification 4.4.6.0.
The portion of the tube bridged by the sleeve joints is effectively
removed from the pressure boundary, and the sleeve then forms the
new pressure boundary. The areas of the sleeved tube assembly which
require inspection are defined in WCAP-13088, Revision 4.
The effect of sleeving on the design transients and accident
analyses have been reviewed based on the installation of sleeves up
to the level of steam generator tube plugging coincident with the
minimum reactor flow rate. The installation of sleeves is to be
evaluated as the equivalent of some level of steam generator tube
plugging. Evaluation of the installation of sleeves is based on the
determination that LOCA evaluations for the licensed minimum reactor
coolant flow bound the effect of a combination of tube plugging and
sleeving up to an equivalent of the actual steam generator tube
plugging limit. Information provided in WCAP-13088, Revision 4,
describes the method to determine the flow equivalency for all
combinations of tubesheet and tube support plate sleeves in order
that the minimum flow requirements are met.
Implementation of laser welded sleeving will reduce the
potential for primary-to-secondary leakage during a postulated steam
line break while maintaining available primary coolant flow area in
the event of a LOCA. By effectively isolating degraded areas of the
tube through repair, primary pressure boundary integrity is restored
and the potential for primary-to-secondary leakage during all plant
conditions is minimized. These degraded tubes are returned to a
condition consistent with the design basis. While the installation
of a sleeve causes a reduction in primary coolant flow, the
reduction is significantly below the reduction incurred by plugging.
Therefore, greater primary coolant flow area is maintained through
sleeving.
Removal of the references to the interim use of an L* repair
criteria will not involve a significant reduction in margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: January 10, 1997
Description of amendment request: The proposed amendment would
modify the Watts Bar Nuclear Plant (WBN) Unit 1 Technical
Specifications (TS) in order to implement the 1995 rule change to 10
CFR Part 50, Appendix J. The revised Appendix J provided an Option B
which allows performance based testing for containment leakage rate
testing. The TS in Section 3.6 and associated Bases, TS Section 3.0.2
and TS Section 5.7 would be changed. Also, the schedular exemption for
containment airlock testing now specified in the facility license in
Section 2.D(1) would no longer be required and would be deleted.
Basis for proposed no significant hazards 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 or consequences of an accident previously evaluated.
The proposed amendment to WBN TSs is in accordance with Option B
to 10 CFR 50, Appendix J. The proposed amendment adds a voluntary
performance-based option for containment leak-rate testing. The
changes being proposed do not affect the precursor for an accident
or transient analyzed in Chapter 15 of WBN Final Safety Analysis
Report. The proposed change does not increase the total allowable
primary containment leakage rate. The proposed change does not
reflect a revision to the physical design and/or operation of the
plant. [T]herefore, operation of the facility, in accordance with
the proposed change, does not significantly affect the probability
or consequences of an accident previously evaluated.
(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 proposed amendment to WBN TSs is in accordance with the new
performance-based option (Option B) to 10 CFR 50,
[[Page 4357]]
Appendix J. The changes being proposed will not change the physical
plant or the modes of operation defined in the facility license. The
proposed changes do not increase the total allowable primary
containment leakage rate. The changes do not involve the addition or
modification of equipment, nor do they alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the propsoed change does not create the possibility of a new or
diferent kind of accident from any previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in margin of
safety.
The proposed change to WBN TSs is in accordance with the new
option to 10 CFR 50, Appendix J. The proposed option is formulated
to adopt performance-based approaches. This option removes the
current prescriptive details from the TS. The proposed changes do
not affect plant safety analyses or change the physical design or
operation of the plant. The proposed change does not increase the
total allowable primary containment leakage rate. Therefore,
operation of the facility, in accordance with the proposed change,
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: November 26, 1996, as supplemented
December 17, 1996
Description of amendment request: The proposed amendments would
allow a one-time only change necessary to replace the existing 125-volt
dc battery cells with new cells. Date of publication of individual
notice in Federal Register: December 13, 1996 (61 FR 65605)
Expiration date of individual notice: January 13, 1997
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina 28223-0001
Notice Of Issuance Of Amendments To Facility Operating LIcenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: August 23, 1996
Brief description of amendment: This amendment makes Technical
Specifications changes allowing fuel enrichment of up to 5.0 weight
percent Uranium-235. The previous limit was 4.1 weight percent. This
change allows Arkansas Nuclear One, Unit-2, to receive, store, and use
nuclear fuel of 5.0 weight percent Uraninum-235.
Date of issuance: January 14, 1997
Effective date: January 14, 1997
Amendment No.: 178
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52964) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 14, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of application for amendment: November 24, 1996, as
supplemented on December 2, 1996.
Brief description of amendment: This amendment adds small break-
loss-of coolant accident methodology CENPD-137, Supplement 1-P and its
approval letter by the NRC as a reference to Section 6.9.5.1. This code
previously approved by the NRC increases the steam generator tube
plugging limit to 30% with an associated reduction of 10% in RCS flow.
This amendment also corrects a typographical error in Specification
6.9.5.1.8, and Specifications 6.9.5.1.10 through 6.9.5.1.14 are
numbered to accommodate these changes.
Date of issuance: January 14, 1997
Effective date: January 14, 1997
Amendment No.: 179
Facility Operating License No. NPF-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 3, 1996 (61 FR
64173) However, on December 9, 1996,
[[Page 4358]]
the licensee verified that the number of plugged tubes would not exceed
their current 10% limit established by the old code. This determination
removed the basis for considering this request as exigent. Since the
potential does exist for the plugging to exceed the 10% in the future,
the technical specification amendment request is therefore, a valid
request on a normal schedule. This change did not alter the staff's
initial proposed no safety hazard condition determination, therefore
noticing was not warranted. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated January 14, 1997.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 24, 1996
Brief description of amendment: The amendment revises the technical
specifications to delete the accelerated testing requirements for the
standby diesel generators. This action is consistent with the
provisions of Generic Letter 94-01, ``Removal of Accelerated Testing
and Special Reporting Requiremets for Emergency Diesel Generators,''
dated May 31, 1994.
Date of issuance: January 14, 1997
Effective date: January 14, 1997
Amendment No.: 90
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications/operating license.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64384) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 14, 1997.No significant
hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 30, 1996
Brief description of amendment: The amendment revises the technical
specification surveillance requirement 3.8.3.4 to specify a 5-start
pressure for the air recievers associated with the Division III, High
Pressure Core Spray emergency diesel generator.
Date of issuance: January 16, 1997
Effective date: January 16, 1997
Amendment No.: 91
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34892) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 16, 1997.No significant
hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: September 23, 1996
Brief description of amendment: Changes to Technical Specification
(TS) to delete a note for the Surveillance Requirement 3.3.7.1 for the
Engineered Safeguard Actuation System Logic.Date of issuance: January
6, 1997
Effective date: January 6, 1997
Amendment No.: 155
Facility Operating License No. DPR-72. Amendment revised the TS.
Date of initial notice in Federal Register: October 23, 1966 (61 FR
55034) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 6, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
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: October 23, 1996, as supplemented by
letter dated November 6, 1996.
Brief description of amendments: The amendments revised Technical
Specification 3.4.6.1, regarding reactor coolant system leakage
detection instrumentation, to adopt the requirements found in NUREG-
1431, ``Standard Technical Specifications Westinghouse Plants,'' for
the reactor coolant system leakage detection instrumentation.
Date of issuance: January 8, 1997
Effective date: January 8, 1997
Amendment Nos.: Unit 1 - Amendment No. 86; Unit 2 - Amendment No.
73
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64387) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 8, 1997.No significant
hazards consideration comments received: No
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
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: April 19, 1996, and as
supplemented on August 15, 1996
Brief description of amendment: The amendment introduces new
Technical Specification (TS) 3.10.10, ``Single Control Rod Withdrawal -
Refueling,'' under TS 3.10, ``SPECIAL OPERATIONS.'' The purpose of this
Special Operations LCO is to permit the withdrawal of a single control
rod for testing in MODE 5 without imposing the requirements for
establishing the secondary containment and main control room boundaries
as normally required during Core Alterations.
Date of issuance: January 13, 1997
Effective date: January 13, 1997
Amendment No.: 112
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 22, 1996 (61 FR
25707) and September 25, 1996 (61 FR 50344). The August 15, 1996,
submittal changed the focus of the original amendment request,
therefore, it was re-noticed in the FEDERAL REGISTER. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated January 13, 1997.No significant hazards consideration comments
received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: September 12, 1995
[[Page 4359]]
Brief description of amendment: The amendment revises Technical
Specification 6.3.1 to add a requirement that the Assistant Operations
Manager hold a senior reactor operator (SRO) license if the Operations
Manager does not hold an SRO license for Millstone Unit 3.
Date of issuance: January 7, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 132
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13530) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 7, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: April 26, 1996, as supplemented
August 23, 1996.
Brief description of amendment: The amendment changes requirements
regarding reactor coolant system leakage testing following refueling
outage and other sytem pressure testing of reactor coolant system
following repairs, replacements, or modifications.
Date of issuance: January 7, 1997
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 171
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28602) The August 23, 1996, letter provided clarifying information that
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 January 7, 1997No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: October 29, 1996
Brief description of amendment: The amendment corrects an error
with respect to Table 3.3.2-1, Function 6c of the Technical
Specifications (TSs) which references the incorrect Required Action for
inoperable channels of the auxiliary feedwater pump actuation on Steam
Generator Level - Low Low logic. The TSs are revised to correct the
Required Action to place the inoperable channel in ``trip'' within 6
hours or initiate a plant shutdown to Mode 4.
Date of issuance: January 9, 1997
Effective date: January 9, 1997
Amendment No.: 66
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64395) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 9, 1997No significant
hazards consideration comments received: No
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of application for amendment: October 29, 1996
Brief description of amendment: This amendment revises the MODE of
applicability for the motor-driven auxiliary feedwater pump actuation
on opening of the main feedwater pump breakers to correct an error
introduced during Amendment No. 61.
Date of issuance: January 9, 1997
Effective date: January 9, 1997
Amendment No.: 67
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64395) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 9, 1997No significant
hazards consideration comments received: No
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: April 29, 1996, as supplemented
October 21, December 2, and December 16, 1996
Brief description of amendments: These amendments revise Technical
Specification (TS) Section 15.3.14, ``Fire Protection System,'' and
Section 15.4.15, ``Fire Protection System,'' and relocate the
requirements of the fire protection program from the TS and
incorporate, by reference, the NRC-approved fire protection program
into the Final Safety Analysis Report. In addition, the amendments
revise the operating licenses to include the NRC's standard fire
protection condition. The amendments also approve administrative
changes consistent with the relocation as well as corrections to
several typographical errors.
Date of issuance: January 8, 1997
Effective date: January 8, 1997, and implementation within 90 days
from the date of issuance. Implementation shall include the relocation
of Technical Specification requirements to the appropriate licensee-
controlled document as identified in the licensee's application dated
April 29, 1996, as supplemented October 21, December 2, and December
16, 1996, and reviewed in the staff's safety evaluation dated January
8, 1997.
Amendment Nos.: Unit 1 - 170, Unit 2 - 174
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revise the Technical Specifications and the operating licenses.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28621) The October 21, December 2, and December 16, 1996, supplements
provided corrected license and TS pages and a 90-day implementation
schedule. These supplements were within the scope of the original
application and did not change the staff's initial proposed no
significant hazards considerations determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated January 8, 1997.No significant hazards consideration
comments received: No
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: October 31, 1996
[[Page 4360]]
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant Technical Specification 6.9, ``Reporting
Requirements,'' by deleting the annual requirement to submit a
description of changes made pursuant to 10 CFR 50.59. Administrative
changes are also made to correct inconsistencies in the TS Table of
Contents and in a footnote for Table TS 3.5-1.
Date of issuance: January 6, 1997
Effective date: January 6, 1997
Amendment No.: 131
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64397) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 6, 1997.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Dated at Rockville, Maryland, this 22nd day of January 1997.
For The Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects - III/IV, Office of
Nuclear Reactor Regulation
[Doc. 97-1994 Filed 1-28-97; 8:45 am]
BILLING CODE 7590-01-F