Biweekly Notice
Federal RegisterJun 21, 1995
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[Federal Register Volume 60, Number 119 (Wednesday, June 21, 1995)]
[Notices]
[Pages 32359-32381]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-50621]
[[Page 32359]]
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 May 26, 1995, through June 9, 1995. The last
biweekly notice was published on Tuesday, June 6, 1995 (60 FR 29869).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By July 21, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any [[Page 32360]] limitations in the order granting leave
to intervene, and have the opportunity to participate fully in the
conduct of the hearing, including the opportunity to present evidence
and cross-examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: June 8, 1995, supersedes December 16,
1994, request in its entirety, supplemented by letters dated November
30, 1994, April 27, 1995, May 5 and May 11, 1995.
Description of amendment request: The proposed amendment would
revise Figure 3.4-4a in the Braidwood Unit 1's technical specifications
which provides the nominal pressurizer power operated relief valve set
points for the low-temperature overpressure protection system (LTOPS).
The proposed revision would extend the applicability of Figure 3.4-4a
from 5.37 effective full power years (EFPY) to 16 EFPY (Unit 1). In
addition, the proposed amendment removes the 638 psig administrative
limit line from the LTOPS curve, because the appropriate instrument
uncertainties and discharge piping pressure limits are included in the
proposed LTOPS curve. The amendment request also proposes
administrative changes to Figure 3.4-4a format and its associated index
page.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The new LTOPS curve will not change any postulated accident
scenarios. The revised curve was developed using industry standards
and regulations which are recognized as being inherently
conservative. Appropriate instrument uncertainties and allowances
have been included in the development of the LTOPS curves. The PT
and LTOPS curves provide RCS pressure limits to protect the Reactor
Pressure Vessel (RPV) from brittle fracture by clearly separating
the region of normal operations from the region where the RPV is
subject to brittle fracture.
Using Regulatory Guide (RG) 1.99, ``Radiation Embrittlement of
Reactor Vessel Materials,'' Revision 2, Braidwood Unit 1
Surveillance Capsule U and Capsule X results and the requirements of
Appendix G to 10 CFR 50, as modified by the guidance in ASME Code
Case N-514, a new LTOPS curve was prepared. This new curve, in
conjunction with the PT Limit curves, and the heatup and cooldown
ranges provides the required assurance that the RPV is protected
from brittle fracture.
No changes to the design of the facility have been made, no new
equipment has been installed, and no existing equipment has been
removed or modified. This amendment will not change any system
operating modes. The revised LTOPS curve provides assurance that the
RPV is protected from brittle fracture.
The index page and format changes are purely administrative in
nature and are designed to reflect the change in the duration of
applicability of Figure 3.4-4a and improve the readability of Figure
3.4-4a. These administrative changes will have no effect on any
equipment, system, or operating mode.
Thus, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The use of the new LTOPS curve does not change any postulated
accident scenarios. The new LTOPS curve was generated using
Braidwood capsule surveillance data and an approved, conservative
methodology. No new equipment will be installed, and no existing
equipment will be modified. No new system interfaces are created,
and no existing system interfaces are modified. The new LTOPS curve
provides assurance that the RPV is protected from brittle fracture.
No new accident or malfunction mechanism is introduced by this
amendment.
The index page and format changes are purely administrative in
nature and are designed to reflect the change in the duration of
applicability of Figure 3.4-4a, and improve the readability of
Figure 3.4-4a. These administrative changes will have no effect on
any equipment, system, or operating mode.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The new LTOPS curve was developed using industry standards and
regulations which are recognized as being inherently conservative.
Appropriate instrument uncertainties and allowances are included in
the development of the new LTOPS curve. This amendment will not
change the operational characteristics or design of any equipment or
system.
All accident analysis assumptions and conditions will continue
to be met. The RPV is adequately protected from non-ductile failure
by the revised LTOPS curve.
The index page and format changes are purely administrative in
nature and are designed to reflect the change in the duration of
applicability of Figure 3.4-4a, and improve the readability of
Figure 3.4-4a. These administrative changes will have no effect on
any equipment, system, or operating mode.
Thus, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this [[Page 32361]] 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: Wilmington Public Library, 201
S. Kankakee Street, Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of amendment request: March 4, 1993, as revised April 14,
1993, as supplemented April 19 and May 31, 1995
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to conform to the wording of
the revised 10 CFR Part 20, ``Standards for Protection Against
Radiation,'' and to reflect a separation of chemistry and radiation
protection responsibilities. The supplemental submittals provided
additional information on the proposed TS change in response to NRC's
request for additional information of May 5, 1995. The original
submittal was noticed on May 12, 1993 (58 FR 28053), as corrected June
1, 1993 (58 FR 31222).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1.Will the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not affect the probability or
consequences of an accident. The proposed change is to the
ADMINISTRATIVE and RADIOLOGICAL EFFLUENT RELEASES sections of the
facility Technical Specifications, and are administrative in nature.
- Change ``Chemistry and Radiation Protection Supervisor'' to
``Radiation Protection Supervisor.''
- The change from ``mR/h'' to ``mrem/h'' is solely a change in
terminology since the revised 10 CFR 20 does not recognize or define
the roentgen as a unit of radiation.
- The Liquid Effluents Concentration section and the associated
bases have been revised to conform with 10 CFR 50.36(a) [10 CFR
50.36a] with effluent concentrations limited to 10 times the limits
of 10 CFR 20.1001 - 20.2402, Appendix B, Table 2, Column 2.
- The actual instantaneous dose rate limits of the Gaseous
Effluents Dose Rate section have not changed. However, the bases
section has. Under the former 10 CFR 20, these dose rates correspond
roughly to maximum permissible concentration and dose(s) received by
the maximum exposed member of the public if allowed to continue for
an entire year. These limits are used more as instantaneous limits
(dose rates above which are not allowed to continue for more than
one hour at a time) so as to provide assurance not to exceed 10 CFR
50, Appendix I limits.
2. Will the proposed change(s) create the possibility of a new
or different kind of accident from any accident previously
evaluated?
This proposed change is required by the implementation of a new
10 CFR Part 20 requirements (except for the title change) and are
administrative in nature (sic). Neither the material condition of
the facility nor the accident analyses are affected by this proposed
change. Therefore, the proposed change does not create the
possibility of a different type of accident than previously
evaluated.
3. Will the proposed change involve a significant reduction in
the margin of safety?
Each limit that was affected increased the margin of safety by
making the limit more conservative; or remained the same.
- The change of distance to ``30 centimeters'' (12 inches) is
more conservative, providing a higher degree of protection for
occupationally exposed worker.
- The liquid effluent concentration limits remain essentially
the same. The bases have changed to [10 CFR 50.36a] reflect 10 times
10 CFR 20.1001 - 20.2402, Appendix B, Table 2, Column 2 limits as
controlled by 10 CFR 50.36(a) [10 CFR 50.36a] dose limits.
- Effluent alarm setpoints were reviewed to determine any
necessary changes and were found to be set appropriately. No change
will be necessary.
- ``The instantaneous release rate limits for airborne releases
will not be changed because they are imposed on licensees as a
control to ensure that the licensees meet Appendix I requirements.''
Alarm setpoints for these dose rate limits may change slightly due
to changes in scientific data and will be reviewed and changed as
appropriate prior to implementation.
Therefore, the proposed change does not involve a reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: Cynthia A. Carpenter, Acting
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 12, 1995
Description of amendment request: The amendments delete Technical
Specification 3/4.3.4, ``Turbine Overspeed Protection,'' and its
associated Bases. The deletion of TS 3/4.3.4 and its associated Bases
provides Duke Power Company the flexibility to implement the
manufacturer's recommendations for turbine steam valve surveillance
test requirements. These test requirements will be relocated from the
TS to the Selected Licensee Commitments (SLC) Manual. The SLC Manual is
Chapter 16 of the Updated Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Relocation of the affected TS section to the SLC Manual
will have no effect on the probability of any accident occurring. In
addition, the consequences of an accident will not be impacted since
the above system will continue to be utilized in the same manner as
before. No impact on the plant response to accidents will be
created.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No new accident causal mechanisms will be created as a
result of relocating the affected TS requirements to the SLC Manual.
Plant operation will not be affected by the proposed amendments and
no new failure modes will be created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. No impact upon any plant safety
margins will be created. Relocation of the affected TS requirements
to the SLC Manual in consistent with the content of the Westinghouse
RSTS [Revised Standard Technical Specifications], as the NRC did not
require technical specification controls for the turbine overspeed
protection system in the RSTS. The proposed amendments are
consistent with the NRC philosophy of encouraging utilities to
propose amendments that are consistent with the content of the RSTS.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this [[Page 32362]] review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 18, 1995, as supplemented by letter
dated May 31, 1995.
Description of amendment request: The proposed amendment would
change Tecnical Specification (TS) 3.6.1.2 to defer the next scheduled
containment integrated leak rate test (ILRT) at Catawba, Unit 2, for
one outage, from the end-of-cycle (EOC) 7 refueling outage (scheduled
for October 1995) to EOC-8 (scheduled for March 1997). Title 10 of the
Code of Federal Regulations, Part 50, Appendix J, requires that three
ILRTs be performed at approximately equal intervals during each 10-year
service period at a nuclear station. ``Approximately equal intervals''
is defined in Catawba's TS as 40 plus or minus 10 months. The proposed
one-time change would allow Catawba to extend that interval to less
than or equal to 70 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Containment leak rate testing is not an initiator of any
accident; the proposed interval extension does not affect reactor
operations or accident analysis, and has no perceptible radiological
consequences. Therefore, this proposed change will not involve a
significant increase in the probability or consequences of any
previously[]evaluated accident.
2. The proposed change will not create the possibility of any
new accident not previously evaluated.
The proposed change does not affect normal plant operations or
configuration, nor does it affect leak rate test methods. The test
history at Catawba (no ILRT [intergrated leak rate test] failures)
provides continued assurance of the leak tightness of the
containment structure.
3. There is no significant reduction in a margin of safety.
It has been documented in draft NUREG-1493 that an increase in
the ILRT interval from 1 test every 3 years to 1 test every 10 years
would result in an increase in population exposure risk in the
vicinity of 5 representative plants from .02% to .14%. The proposed
change included herein, an increase from 40 [plus or minus] 10
months to [less than or equal to] 70 months, represents a small
fraction of that already very small increase in risk. Therefore, it
may be concluded that no significant reduction in a margin of safety
will occur.
Based on the above, no significant hazards consideration is
created by the proposed 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
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: April 12, 1995
Description of amendment request: The amendments delete Technical
Specification 3/4.3.4, ``Turbine Overspeed Protection,'' and its
associated Bases. The deletion of TS 3/4.3.4 and its associated Bases
provides Duke Power Company the flexibility to implement the
manufacturer's recommendations for turbine steam valve surveillance
test requirements. These test requirements will be relocated from the
TS to the Selected Licensee Commitments (SLC) Manual. The SLC Manual is
Chapter 16 of the Updated Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Relocation of the affected TS section to the SLC Manual
will have no effect on the probability of any accident occurring. In
addition, the consequences of an accident will not be impacted since
the above system will continue to be utilized in the same manner as
before. No impact on the plant response to accidents will be
created.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No new accident causal mechanisms will be created as a
result of relocating the affected TS requirements to the SLC Manual.
Plant operation will not be affected by the proposed amendments and
no new failure modes will be created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. No impact upon any plant safety
margins will be created. Relocation of the affected TS requirements
to the SLC Manual in consistent with the content of the Westinghouse
RSTS [Revised Standard Technical Specifications], as the NRC did not
require technical specification controls for the turbine overspeed
protection system in the RSTS. The proposed amendments are
consistent with the NRC philosophy of encouraging utilities to
propose amendments that are consistent with the content of the RSTS.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do 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: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Florida Power and Light Company, et al., Docket No. 50-335, St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: May 17, 1995
Description of amendment request: The amendment will extend the
applicability of the current Reactor Coolant System (RCS) Pressure/
Temperature Limits and maximum allowed RCS heatup and cooldown rates to
23.6 Effective Full Power Years (EFPY) of operation. In addition,
administrative changes are proposed for [[Page 32363]] TS 3.1.2.1
(Boration Systems Flow Paths-Shutdown) and TS 3.1.2.3 (Charging Pump-
Shutdown) to clarify the conditions for which a High Pressure Safety
Injection pump may be used.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if 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. Each standard is discussed as
follows:
(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 pressure-temperature (P/T) limit curves in the Technical
Specifications are conservatively generated in accordance with the
fracture toughness requirements of 10 CFR 50 Appendix G as
supplemented by the ASME Code Section XI, Appendix G
recommendations. The RTNDT values are based on Regulatory Guide
1.99, Revision 2, shift prediction and attenuation formula. Analyses
of reactor vessel material irradiation surveillance specimens are
used to verify the validity of the fluence predictions and the P/T
limit curves. Use of these curves in conjunction with the
surveillance specimen program ensures that the reactor coolant
pressure boundary will behave in a non-brittle manner and that the
possibility of rapidly propagating fracture is minimized. Based on
the use of plant specific material data, analysis has demonstrated
that the current P/T limit curves will remain conservative for up to
23.6 EFPY.
In conjunction with extending the applicability of the existing
P/T limit curves, the low temperature overpressure protection (LTOP)
analysis for 15 EFPY is also extended. The LTOP analysis confirms
that the current setpoints for the power-operated relief valves
(PORVs) will provide the appropriate overpressure protection at low
Reactor Coolant System (RCS) temperatures. Because the P/T limit
curves have not changed, the existing LTOP values have not changed,
which include the PORV setpoints, heatup and cooldown rates, and
disabling of non-essential components.
The proposed amendment does not change the configuration or
operation of the plant, and assurance is provided that reactor
vessel integrity will be maintained. Therefore, operation of the
facility in accordance with the proposed amendment will not involve
a significant increase in 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.
By applying plant specific data in the determination of critical
vessel material limits, the applicability of the existing pressure
temperature limits and LTOP requirements can be extended. There is
no change in the configuration or operation of the facility as a
result of the proposed amendment. The amendment does not involve the
addition of new equipment or the modification of existing equipment,
nor does it alter the design of St. Lucie plant systems. Therefore,
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.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Analysis has demonstrated that the fracture toughness
requirements of 10 CFR 50 Appendix G are satisfied and that
conservative operating restrictions are maintained for the purpose
of low temperature overpressure protection. The P/T limit curves
will provide assurance that the RCS pressure boundary will behave in
a ductile manner and that the probability of a rapidly propagating
fracture is minimized. Therefore, operation of the facility in
accordance with the proposed amendment would not involve a
significant reduction in a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this
proposed license amendment involves no 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: May 17, 1995
Description of amendment request: The proposed amendments will
improve consistency between the Technical Specifications and the
improved Combustion Engineering Standard Technical Specifications
(NUREG-1432, dated September 1992) by incorporating changes in text and
resolving other inconsistencies identified by the NRC and plant
operations staff.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if 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. Each standard is discussed as
follows:
(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 amendments consist of administrative changes to the
Technical Specifications (TS) for St. Lucie Units 1 and 2. The
amendments will implement changes in text to improve consistency
within the TS for each unit, the improved Combustion Engineering
Standard Technical Specifications (NUREG-1432, dated September
1992), and the regulations. The proposed amendments do not involve
changes to the configuration or method of operation of plant
equipment that is used to mitigate the consequences of an accident,
nor do the changes otherwise affect the initial conditions or
conservatism assumed in any of the plant accident analyses.
Therefore, operation of the facility in accordance with the proposed
amendments would not involve a significant increase in 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 administrative revisions will not change the
physical plant or the modes of plant operation defined in the
Facility License for each unit. 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 proposed amendments would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not [[Page 32364]] involve a significant reduction
in a margin of safety.
The proposed amendments are administrative in nature and do not
change the basis for any technical specification that is related to
the establishment of, or the preservation of, a nuclear safety
margin. Therefore, operation of the facility in accordance with the
proposed amendments would not involve a significant reduction in a
margin of safety.
Based on the above discussion and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: May 23, 1995
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) by changing
the setpoint presentation format for the Reactor Protection System
(RPS) and Engineered Safety Features Actuation System (ESFAS)
instrumentation setpoints contained in Technical Specification Tables
2.2-1 and 3.3-3. The approved Westinghouse five-column instrument
setpoint methodology currently being used to establishing those
setpoints would be retained. The intent of the amendments is to
eliminate the need for minor administrative license amendments to these
tables that do not impact either the Trip Setpoints or the Safety
Analysis Limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
No changes to the Reactor Trip System instrumentation setpoints,
ESFAS instrumentation setpoints, or the Turkey Point Plant licensing
basis (NRC-approved, Westinghouse five-column setpoint methodology,
as documented in Westinghouse topical report WCAP-12745P), is being
made. The changes proposed reduce the level of detail in the
Technical Specifications and place that detailed information in
controlled procedures, drawings and the Final Safety Analysis
Report. Since the setpoints and methodology remain the same, the
changes proposed by this submittal will not increase the probability
or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
These proposed changes remove from the Technical Specifications
a level of detail which will be maintained in controlled procedures
and drawings. The Turkey Point Plant licensing basis (NRC-approved,
Westinghouse five column setpoint methodology, as documented in
Westinghouse topical report WCAP-12745P), continues to be used to
calculate the Reactor Trip System and ESFAS setpoints. No changes to
Reactor Trip System or ESFAS instrumentation setpoints are proposed.
Since the same methodology will be used to determine the setpoints
and no setpoints are changed, the possibility that a new or
different kind of accident from any previously evaluated will not be
created.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The Turkey Point Plant licensing basis (NRC-approved,
Westinghouse five column setpoint methodology, as documented in
Westinghouse topical report WCAP-12745P), continues to be used to
calculate the Reactor Trip System and ESFAS setpoints. No changes to
the Reactor Trip System or ESFAS instrumentation setpoints are
proposed. Since the same methodology will be used to determine the
setpoints, and no setpoints are changed by this submittal, this
change does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: David B. Matthews
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: June 6, 1995
Description of amendment request: The proposed change would revise
Plant Hatch Units 1 and 2 Technical Specification (TS) Surveillance
Requirements (SR) 3.6.4.1.3 and 3.6.4.1.4 for the secondary containment
drawdown. The revision would reduce the SR acceptance criteria to
greater than or equal to 0.20 inch of vacuum from greater than or equal
to 0.25 inch of vacuum. Also, the licensee proposed to change the Bases
to reflect the proposed TS revision.
The licensee stated that the secondary containment performs no
active function in response to either loss-of-coolant accident or fuel
handling accident. However, its leak tightness is required to ensure
that the release of radioactive materials from the primary containment
is restricted to those leakage paths and associated leakage rates
assumed in the accident analysis and that fission products entrapped
within the secondary containment structure will be treated by the Unit
1 and Unit 2 standby gas treatment systems prior to discharge to the
environment. This change will continue to provide adequate margin for
the secondary containment to be sufficiently leak tight such that the
conclusions of the accident analysis remain valid.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
secondary containment serves a mitigation function and therefore
this change does not increase the probability of an accident
previously evaluated. The consequences of the previously evaluated
accidents are not affected because at the wind conditions assumed in
the accident analysis the building will be at a negative pressure
and no exfiltration is postulated. Furthermore, the estimated wind
speed at which exfiltration might take place (31 mph) is not a
frequent occurrence (wind speeds of greater than 24 mph occur [less
than] NDT of 22 deg.F (i.e., 70 deg. - 48 deg. = 22 deg.).
This is a ``margin'' over and above the boltup temperature required
by Appendix G to the 1992 ASME Section XI Code, since Appendix G
would allow a boltup temperature of 48F.
The above proposed changes to the Limiting Condition for
Operation for tensioning the reactor vessel head studs do not alter
the configuration, normal operation, design bases, function,
mission, or performance of the subject components. Therefore, the
proposed changes do not affect the margin of safety inherent in the
design, analysis, function, or operation of the reactor vessel head
flange region. The proposed changes do not alter the fuel clad
barrier, fuel integrity, reactor vessel integrity, reactor coolant
system integrity, or the containment boundary integrity; thus the
margin of safety related to these barriers remains unchanged.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: May 24, 1995
Description of amendment request: The proposed amendment would
permit an individual who does not have a current senior reactor
operator (SRO) license to hold the Operations Manager position. The
position will require the individual to have previously held an SRO
license at a boiling water reactor (BWR). An individual serving in the
capacity of the Assistant Operations Manager will hold a current SRO
license for Millstone Unit 1, if the Operations Manager does not. In
addition, the proposed amendment would renumber the applicable sections
of the related technical specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed change in accordance with
10CFR50.92 and concluded that the change does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed change does not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed change affects an administrative control, which was
based on the guidance of ANSI N18.1-1971. ANSI N18.1-1971
recommended that the Operations Manager hold an SRO license. The
current guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends,
as one option, that the Operations Manager have held a license for a
similar unit and the Operations Middle Manager hold an SRO license.
While the Operations Middle Manager position does not exist at
Millstone Unit No. 1, NNECO has created the position of Assistant
Operations Manager. The individual in this position would meet the
requirements for, and would have responsibilities as recommended in,
ANSI/ANS 3.1-1987 for the Operations Middle Manager position.
Therefore, the proposed change requests an exception to ANSI
N18.1-1971 to allow use of ANSI/ANS 3.1-1987 in a limited
circumstance. Specifically, the proposed revision to Technical
Specification 6.3.1 would require the Operations Manager to either
hold an SRO license at Millstone Unit No. 1 or have held an SRO at a
BWR.
If the Operations Manager does not hold an SRO license at
Millstone Unit No. 1, the specification will require the Assistant
Operations Manager to hold, and continue to hold, an SRO license.
The proposed change includes the requirement to have held a license
for a similar unit (a BWR) in accordance with Section 4.2.2 of ANSI/
ANS 3.1-1987, if the Operations Manager does not hold an SRO license
at Millstone Unit No. 1. For those areas of knowledge that require
an SRO license, the Assistant Operations Manager will provide the
technical guidance typically provided by the Operations Manager.
The proposed change does not alter the design of any system,
structure, or component, nor does it change the way plant systems
are operated. It does not reduce the knowledge, qualifications, or
skills of licensed operators, and does not affect the way the
Operations Department is managed by the Operations Manager. The
Operations Manager will continue to maintain the effective
performance of his personnel and ensure the plant is operated safely
and in accordance with the requirements of the operating license.
Additionally, the control room operators will continue to be
supervised by the licensed Shift Supervisor.
The proposed change does not detract from the Operations
Manager's ability to perform his primary responsibilities. In this
case, by having previously held an SRO license, the Operations
Manager has achieved the necessary training, skills, and experience
to fully understand the operation of plant equipment and the watch
requirements for operators. In summary, the proposed change does not
affect the ability of the Operations Manager to provide the plant
oversight required of that position. Thus, it does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change to Technical Specification 6.3.1 does not
affect the design or function of any plant system, structure, or
component, nor does it change the way plant systems are operated. It
does not affect the performance of licensed operators. Operation of
the plant in conformance with technical specifications and other
license requirements will continue to be supervised by personnel who
hold an SRO license. The proposed change to Technical Specification
6.3.1 ensures that the Operation Manager will be a knowledgeable and
qualified individual by requiring the individual to have held an SRO
license at a BWR. Based on the above, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed change involves an administrative control that is
not related to the margin of safety. The proposed change does not
reduce the level of knowledge or experience required of an
individual who fills the Operations Manager position, nor does it
affect the conservative manner in which the plant is operated. The
Control Room operators will continue to be supervised by personnel
who hold an SRO license. Thus, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 32371]] amendment request involves no significant hazards
consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: May 26, 1995
Description of amendment request: The proposed amendment will
delete the old limiting conditions for operation (LCOs) and
surveillance requirements and add new LCOs, surveillance requirements,
and bases for the loss of normal power (LNP) instrumentation system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed this proposed change in accordance with
10CFR50.92 and concluded that this change does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed change does not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The change does not increase the probability of a loss of off-
site power event or the occurrence of any accidents which assume
loss of off-site power. This is ensured by the LNP instrumentation
system design which uses multiple sensing relays, redundancy, and
qualified Class 1E components, as well as conservative operability
and surveillance requirements.
Full LNP logic requires two sets of relays to trip in one of two
redundant groups. One set monitors bus 14E and the other set
monitors bus 14F. Separate sets are provided for loss of voltage and
degraded voltage monitoring. This design minimizes the likelihood of
an inadvertent full LNP initiation. To maintain redundancy in the
instrumentation, two separate groups are provided, each group being
powered from an independent DC supply. Partial LNP logic is also
provided to detect a loss of voltage on a single emergency bus.
Redundancy in the partial LNP logic is achieved by providing an
independent logic for each emergency power train.
The proposed technical specification would require that the LNP
instrumentation be maintained operable except when the unit is in
cold shutdown or refueling conditions. If redundancy in the ability
to detect a loss of voltage or degraded voltage and initiate a full
LNP is not maintained, reactor operation would be permitted for
seven days. In this situation, both full and partial LNP (and both
emergency power sources) remain operable. An action statement of
seven days, which is the same as the action statement duration for
an inoperable EDG [emergency diesel generator], is justified based
on continued operability of the other LNP group. Additionally, it
allows a reasonable amount of time to perform repairs.
The time delays and voltage setpoints specified in Table 3.2.4
ensure that the emergency power source starting and loading times
continue to meet the current technical specification requirements.
Also, these time delays are long enough to preclude false trips due
to anticipated voltage transients (e.g., during motor starts). The
relay calibration surveillance procedure will establish acceptance
criteria for each relay to ensure that the total times specified in
Table 3.2.4 are not exceeded. The proposed surveillance testing and
calibration frequency of every refueling outage is consistent with
the requirements in the current technical specification.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
There are no new failure modes associated with this change since
the proposed requirements will ensure the LNP instrumentation system
is available to perform its safety function. Individual voltage
sensing relays, when removed from their cases, would provide the
tripped contact configuration. The proposed technical specification
would allow relays to be placed in the tripped condition as long as
it would not inhibit the LNP function or cause an inadvertent
initiation. Additionally, since the design function to ensure that
adequate power is available to operate the emergency safeguards
equipment has not changed, no new accident or accident of a
different kind is created.
3. Involve a significant reduction in the margin of safety.
The protective boundaries are not affected because the
consequences of any design basis accident are not changed. Since the
protective boundaries are not affected, the safety limits are also
unaffected. The proposed change maintains the basis of the technical
specifications by ensuring that adequate electrical power is
available to operate the emergency safeguards equipment. By
maximizing the operability of the LNP instrumentation without
requiring high risk testing, the proposed change will improve the
margin of safety as related to availability of electric power to
safety related loads.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: January 13, 1995
Description of amendment request: The proposed amendment would
revise the Administrative Controls Section (6.0) of the Technical
Specifications (TS) for Hope Creek Generating Station to reflect
organizational changes and resultant management title changes. As
indicated on the marked-up pages in Attachment 2, PSE&G requests that:
1) Vice President and Chief Nuclear Officer will be replaced with Chief
Nuclear Officer and President - Nuclear Business Unit in TS 6.1.2,
6.2.1.c, 6.5.2.4.3.g, 6.5.2.4.4.a, 6.5.2.4.4.b, 6.5.2.6, 6.6.1.b,
6.7.1.a, and 6.7.1.c. 2) Vice President and Chief Nuclear Officer will
be replaced with Vice President - Nuclear Operations in TS 6.5.1.8.b,
and 6.5.1.9. 3) In addition, General Manager - Quality Assurance and
Nuclear Safety will be replaced with Director - Quality Assurance and
Nuclear Safety Review in TS 6.5.1.8.b, 6.5.1.9, 6.5.2.2, 6.5.2.4.3.g,
6.7.1.a, 6.7.1.c.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed management title changes from Vice President and
Chief Nuclear Officer to Chief Nuclear Officer and President -
Nuclear Business Unit or Vice President - Nuclear Operations, and
from General Manager - Quality Assurance and Nuclear Safety to
Director - Quality Assurance and Nuclear Safety Review are
administrative in nature and do not affect assumptions contained in
the plant safety analysis, the physical design and/or operation of
the plant, nor do they affect Technical Specifications that preserve
safety analysis assumptions. Therefore, the proposed changes do not
[[Page 32372]] involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The changes being proposed are purely administrative and will
not lead to material procedure changes or to physical modifications.
Therefore, the proposed changes do not create the possibility of a
new or different type of accident from any accident previously
evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The changes being proposed are administrative in nature and do
not relate to or modify the safety margins defined in and maintained
by the Technical Specifications. The changes discussed herein do not
reduce the Technical Specification safety margin since all
organizational responsibilities are being adequately implemented,
and all personnel in place are properly qualified. 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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 19, 1995 (TS 95-07)
Description of amendment request: The proposed change would (1)
modify Surveillance Requirement (SR) 4.1.1.3 to allow suspension of the
end of life (EOL) moderator temperature coefficient (MTC) surveillance
measurement provided the benchmark criteria and the Revised Prediction
as documented in the Core Operating Limits Report (COLR) are satisfied.
The SR would also indicate that the data required for the calculation
of the Revised Prediction is provided in the Most Negative Temperature
Coefficient Limit Report per Specification 6.9.1.15. In addition, a
grammatical error affecting the Unit 1 SR would be corrected; (2)
modify Technical Specifications (TS) 6.9.1.14, COLR, by adding to the
list of references: WCAP-13749-P-[A], ``Safety Evaluation Supporting
the Conditional Exemption of the Most Negative EOL Moderator
Temperature Coefficient Measurement,'' May 1993 (Proprietary)
(Methodology for Specification 3.1.1.3 - Moderator Temperature
Coefficient); (3) add Specification 6.9.1.15, which would require that
the Most Negative MTC Report be prepared at least 60 days prior to the
date the limit would become effective and be maintained on file. Also,
the TS would require that the data required for the determination of
the Revised Prediction of the 300 ppm/RTP MTC per WCAP-13749-P-[A] be
included in the report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The conditional exemption of the most negative moderator
temperature coefficient (MTC) measurement does not change the most
negative MTC surveillance requirement (SR) and limiting condition of
operation (LCO) limits in the TSs. Since these MTC values are
unchanged, and since the basis for the derivation of these values
from the safety analysis moderator density coefficient (MDC) is
unchanged, the constant MDC assumed for the Updated Final Safety
Analysis Report (UFSAR) safety analyses will also remain unchanged.
Therefore, no change in the modeling (i.e., probabilities) of the
accident analysis conditions or response is necessary in order to
implement the change to the conditional exemption methodology. In
addition, since the constant MDC assumed in the safety analyses is
not changed by the conditional exemption of the most negative MTC SR
measurement, the consequences of an accident previously evaluated in
the UFSAR are not increased. The dose predictions presented in the
UFSAR for a steam generator tube rupture remain valid such that more
severe consequences will not occur. Additionally, since mass and
energy releases for a loss-of-coolant accident and a steamline break
are not increased as a result of the unchanged MDC, the dose
predictions for these events presented in the UFSAR also remain
bounding.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Since the end-of-life MTC is not changed by the conditional
exemption methodology of WCAP-13749-P, the possibility of an
accident, which is different than any already evaluated in the
UFSAR, has not been created. No new or different failure modes have
been defined for any system or component nor has any new limiting
single failure been identified. Conservative assumptions for the MDC
have already been modeled in the UFSAR analyses. These assumptions
will remain valid since the conditional exemption methodology
documented in WCAP-13749-P does not change the safety analysis MDC
nor the TS values of the MTC.
3. Involve a significant reduction in a margin of safety.
The conditional exemption methodology is documented in WCAP-
13749-P. This WCAP has been evaluated (Reference: SECL 93-117,R1)
relative to the design basis, including the TSs, and has been
determined to bound the conditions under which the specifications
permit operation. The results as presented in the UFSAR remain
bounding since the MDC assumed in the safety analyses and the
limiting conditions for operation and SR MTCs in the TSs remain
unchanged. Therefore, the margin of safety, as defined in the bases
to these TSs, is not reduced.
The NRC has reviewed the licensee's analysis and, based on
thisreview, 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,1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear
Plant, Unit 2, Hamilton County, Tennessee
Date of amendment request: May 19, 1995 (TS 95-13)
Description of amendment request: The proposed change would revise
License Condition 2.C.(17) to extend the required surveillance interval
to May 4, 1996, for Surveillance Requirement 4.3.2.1.3. The proposed
change would extend the Engineered Safety Features Response Time
instrument tests required at 36-month intervals shown in Table 3.3-3
associated with safety injection, feedwater isolation, containment
isolation Phase A, auxiliary feedwater pump, essential raw cooling
water system, emergency gas treatment system, containment spray,
containment isolation Phase B, turbine trip, 6.9-kilovolt shutdown
board-degraded voltage or loss of voltage, and automatic switchover to
containment sump actuations. The proposed extension will limit the
interval past the allowable extension provided by TS 4.0.2 to 4.5
months. [[Page 32373]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c).
Operation of Sequoyah Nuclear Plant (SQN) in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change is temporary and allows a one-time extension
of Surveillance Requirement 4.3.2.1.3 for Cycle 7 to allow
surveillance testing to coincide with the seventh refueling outage.
The proposed surveillance interval extension will not cause a
significant reduction in system reliability nor affect the ability
of the systems to perform their design function. Current monitoring
of plant conditions and continuation of the surveillance testing
required during normal plant operation will continue to be performed
to ensure conformance with TS operability requirements. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Extending the surveillance interval for the performance of
specific testing will not create the possibility of a new or
different kind of accidents. No changes are required to any system
configurations, plant equipment, or analyses. Therefore, this change
will not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
Surveillance interval extensions will not impact any plant
safety analyses since the assumptions used will remain unchanged.
The safety limits assumed in the accident analyses and the design
function of the equipment required to mitigate the consequences of
any postulated accidents will not be changed since only the
surveillance test interval is being extended. Historical performance
generally indicates a high degree of reliability, and surveillance
testing performed during normal plant operation will continue to be
performed to verify proper performance. Therefore, the plant will be
maintained within the analyzed limits, and the proposed extension
will not significantly reduce the margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, 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, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: May 1, 1995
Brief description of amendments: The proposed amendment would: (1)
reduce the minimum fuel oil volume requirement during MODES 5 and 6,
for OPERABLE emergency diesel generators (EDG), and (2) allow continued
OPERABLE status of diesel generators during all MODES, for 48 hours
with greater than 6-day supply of diesel fuel for the associated diesel
generator.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
REDUCTION IN MINIMUM DIESEL FUEL STORED VOLUME WHILE SHUTDOWN
The first proposed change reduces the diesel fuel oil inventory
required during plant shutdown conditions (MODES 5 and 6). The
current fuel oil inventory requirement is the same for plant
operation (MODES 1, 2, 3 and 4) and for plant shutdown. This current
inventory requirement is based upon the seven days continuous
operation of a diesel generator at its rated capacity which
encompasses all load demands for the Loss of Coolant Accident
concurrent with a Loss of Offsite Power (LOCA/LOOP) scenario.
Because of reduced temperature and pressure, LOCA/LOOP is a less
significant and probable event in MODES 5 and 6. The bounding
scenario is considered to be a Loss of Offsite Power (LOOP) while
the plant is shutdown (in MODES 5 and 6). The new diesel fuel oil
inventory required during plant shutdown conditions is based on
LOOP. Because this change only affects diesel fuel inventory, there
is no impact on the probability of an accident. The consequences of
LOOP event are unchanged since sufficient fuel remains available to
allow the diesel generators to support mitigation of the event.
Because seven days of fuel are required, there is no change in the
consequences of any event which requires the diesel generators.
Therefore, there is no significant increase in the probability or
consequences of an accident previously evaluated as a result of this
proposed change.
ADDITION OF REMEDIAL ACTION TO RESTORE THE STORED VOLUME OF
DIESEL FUEL
The second proposed change applies to all MODES of operation.
This change allows the diesel generator to remain OPERABLE if the
fuel oil inventory falls below the minimum required in the storage
system (i.e., fuel volume for 7-day operation of the diesel
generator) but remains above a fuel volume for 6 days operation of
the diesel generator. The minimum required fuel oil volume must be
restored within 48 hours of falling below the limit. This relaxation
by 48 hours allows sufficient time to replenish the required fuel
oil volume and complete any required analysis prior to fuel oil
addition to the storage tank. Because this change only affects
diesel generator fuel inventory, there is no impact on the
probability of an accident. Since the fuel oil replenishment can be
obtained in less than six days after an event, there is no
significant increase in the probability of a loss of all AC power
(i.e., Station Blackout). Because the remaining fuel oil volume is
larger than 6-day fuel supply and actions are initiated to obtain
replenishment within this brief period, the proposed change does not
involve a significant increase in the consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
REDUCTION IN MINIMUM DIESEL FUEL STORED VOLUME WHILE SHUTDOWN
The first proposed change reduces the diesel fuel oil inventory
required for plant shutdown conditions. As described above, LOOP is
the limiting condition for diesel fuel oil inventory requirements
for a plant in the shutdown condition. As the proposed fuel
inventory is adequate for a shutdown LOOP and no hardware changes or
system operation changes are involved, no new failure modes are
introduced and hence, no new or different accidents from any
previously evaluated are created.
ADDITION OF REMEDIAL ACTION TO RESTORE THE STORED VOLUME OF
DIESEL FUEL
The second proposed change only affects diesel generator fuel
inventory as well. There are no hardware changes and no changes in
system operations involved; therefore, no new or different accidents
from any accident previously evaluated are created.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The intent of the Technical Specification is to conservatively
assure sufficient fuel to assure diesel generator operation to
support mitigation of postulated events. This intent is accomplished
by conservatively assuring a seven day supply of fuel. Seven days
fuel supply is considered sufficient to support the initial
mitigation activities, identify the need for additional fuel,
arrange for delivery, test and then add fuel to the storage tanks,
if needed. The current diesel fuel oil inventory for operating
conditions (MODES 1, 2, 3 and 4), is sufficient to conservatively
support seven days of diesel generator operation for a LOCA with
LOOP condition.
REDUCTION IN MINIMUM DIESEL FUEL STORED VOLUME WHILE SHUTDOWN
[[Page 32374]]
The proposed diesel fuel oil inventory for shutdown conditions
(MODES 5 and 6), is adequate to conservatively support seven days of
diesel generator operation for LOOP conditions. The proposed
reduction in inventory between operating and shutdown conditions
continues to support the different transient conditions which are
applicable to the different modes of operation. Even though the
minimum storage requirement during shutdown is being reduced, the
basis of this specification continues to be conservatively satisfied
and therefore this license amendment request does not involve a
significant reduction in a margin of safety.
ADDITION OF REMEDIAL ACTION TO RESTORE THE STORED VOLUME OF
DIESEL FUEL
The second proposed change which is applicable to all MODES of
operation, allows 48 hours to restore diesel generator fuel oil
inventory to the seven-day level as long as the inventory does not
fall below the six-day level. The probability of a LOOP during this
period is low. The 6-day fuel oil supply is calculated with adequate
margin similar to the calculation of 7-day fuel oil inventory. In
spite of the potential that there may be slightly less fuel
available inlenishment within this brief period. Based on this and
the low probability of an event during this brief period, it is
considered that this change request 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: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Newman and
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
NRC Project Director: William D. Beckner
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: May 4, 1994
Brief description of amendments: The amendments revise Limiting
Condition for Operation (LCO) 3.4.8.3 and Surveillance Requirement
4.4.8.3.1, ``Overpressure Protection Systems.'' Specifically, the LCO
and surveillance requirements are revised to clarify that both shutdown
cooling system (SCS) suction relief valves shall be OPERABLE and
aligned to provide overpressure protection not only during reactor
coolant system (RCS) cooldown and heatup evolutions, but also during
any steady-state temperature periods in the course of RCS cooldown or
heatup evolutions.
Date of issuance: June 2, 1995
Effective date: June 2, 1995
Amendment Nos.: Unit 1 - Amendment No. 93; Unit 2 - Amendment No.
80; Unit 3 - Amendment No. 63
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42333) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 2, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, Massachusetts
Date of application for amendment: November 22, 1994
Brief description of amendment: This amendment revises the
suppression chamber water level operating range, increasing it 2
inches, and revises the water level recorder range in response to a
commitment from an inspection.
Date of issuance: June 1, 1995
Effective date: June 1, 1995
Amendment No.: 163
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 1995 (60 FR
3672) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 1, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: February 24, 1995
Brief description of amendment: The proposed change would remove
Section 4.3 from the Technical Specifications (TS) because the primary
system testing following opening is already performed in accordance
with the American Society of Mechanical Engineers Boiler and Pressure
Vessel Code, as implemented in the licensee's inservice inspection
program as required by TS 4.0.1.
Date of issuance: May 30, 1995Effective date: May 30, 1995
Amendment No.: 165
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16183) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 30, 1995.No significant
hazards consideration comments received: No [[Page 32375]]
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois;
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: October 15, 1992, as
supplemented March 9, 1993.
Brief description of amendments: The amendments would modify the
existing Dresden and Quad Cities Technical Specifications (TS) to
format them in the style of the Boiling Water Reactor 4 (BWR) Standard
Technical Specifications (STS). The amendments deal specifically with
Section 3/4.4, ``Standby Liquid Control System (SLCS).''
Date of issuance: June 8, 1995
Effective date: For Dresden, immediately, to be implemented no
later than December 31, 1995; for Quad Cities, immediately, to be
implemented no later than June 30, 1996.
Amendment Nos.: 133, 127, 154, and 150
Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36429) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 8, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: April 10, 1995
Brief description of amendments: The amendments would change the
Technical Specifications by (1) revising the low pressure value at
which the High Pressure Coolant Injection (HPCI) and Reactor Core
Isolation Cooling (RCIC) systems can be tested to 150 psig, and (2)
testing these systems against a system head corresponding to reactor
vessel pressure when steam is supplied to the turbines at 920 psig to
1005 psig for high pressure testing and 150 psig to 325 psig for low
pressure testing.
Date of issuance: May 30, 1995
Effective date: Immediately and shall be implemented within 60
days.
Amendment Nos.: 153 and 149
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1995 (60 FR
21009) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 30, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: November 21, 1994.
Brief description of amendments: The amendments add footnotes in
Limiting Condition for Operation 3.15.2.A of the Technical
Specifications (TS) to allow a one-time extension of the allowed outage
time (AOT) for an inoperable reserve offsite power source from 72 hours
to 14 days. To provide additional assurance that redundant sources of
power to the operating unit are operable during the AOT outage, the
amendment also adds footnotes in Surveillance Requirement 4.15.2.A of
the TS to modify the emergency diesel generator and the normal offsite
power source testing requirements.
Date of issuance: May 31, 1995
Effective date: May 31, 1995
Amendment Nos.: 163 and 151
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
500). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 31, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, MichiganDate of application for amendment: October
20, 1992
Brief description of amendment: This amendment revises Technical
Specification 5.3.1a to account for changes being made to the Palisades
Final Safety Analysis Report (FSAR) Section 4.2 following replacement
of the steam generators.
Date of issuance: May 22, 1995
Effective date: May 22, 1995
Amendment No.: 166
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18624) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 22, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of application for amendment: January 13, 1995, as
supplemented April 12 and 27, 1995
Brief description of amendment: This amendment revises the
Technical Specifications to allow installed primary and secondary
safety valve settings to be within a 3% tolerance of their nominal
settings, but would require returning the valve settings to within 1%
of the nominal settings if the valves are removed from the piping for
maintenance or testing.
Date of issuance: June 8, 1995
Effective date: June 8, 1995
Amendment No.: 167
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11130) The April 12 and 27, 1995, letters provided clarifying
information in response to the staff's request for additional
information of April 11, 1995, and a telephone request for information
on the Palisades loss of load analysis contained in the January 13,
1995, submittal. This information was within the scope of the original
application and did not change the staff's initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
June 8, 1995.No significant hazards consideration comments received:
No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423. [[Page 32376]]
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan Date of application for amendment: September 13, 1993
Brief description of amendment: The amendment revises Technical
Specification (TS) 6.5.2.8 to relocate audit frequencies from the TS to
the Quality Assurance Program located in Chapter 17.2 of the Updated
Final Safety Analysis Report. A related change to extend the frequency
of the use of an independent fire contractor to every third fire
protection audit was denied.
Date of issuance: May 23, 1995
Effective date: May 23, 1995, with full implementation within 45
days.
Amendment No.: 104
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18625) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 23, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 16, 1993
Brief description of amendment: The amendment revised the Technical
Specifications by removing the incore detection system requirements.
These requirements are to be relocated in the Updated Final Safety
Analysis Report.
Date of issuance: May 30, 1995
Effective date: May 30, 1995, to be implemented within 60 days.
Amendment No.: 107
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57851) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 30, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: January 20, 1995
Brief description of amendments: These amendments will relocate the
operability requirements for Incore Detectors in Technical
Specification 3/4.3.3.2 to the Updated Final Safety Analysis Report,
and revise Linear Heat Rate Surveillance 4.2.1.4, and Special Test
Exceptions Surveillances 4.10.2.2, 4.10.4.2 (Unit 2 only), and
4.10.5.2, accordingly.
Date of issuance: June 6, 1995
Effective date: June 6, 1995
Amendment Nos.: 136 and 75
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11132) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 6, 1995No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of application for amendments: December 29, 1994, as
supplemented by letter dated May 2, 1995.
Brief description of amendments: The amendments revise TS 3/4.3,
Instrumentation and its associated Bases, and TS 3/4.8, Electrical
Power Systems to specify the appropriate actions to take in the event
that an automatic load sequencer must be taken out of service or
becomes inoperable.
Date of issuance: May 31, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 86 and 64
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6301). The May 2, 1995, letter provided minor editorial changes that
did not change the scope of the December 29, 1994, application and
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated May 31, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830.
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: January 13, 1993, as supplemented by
letter dated October 18, 1993
Brief description of amendment: The amendment revised the River
Bend Station, Unit 1 operating license to reflect a change in ownership
of Gulf States Utilities (GSU). GSU, which ownes a 70 percent undivided
interest in the River Bend Station, is a wholly-owned subsidiary
company of Entergy Corporation. This amendment was originally issued on
December 16, 1993, as License Amendment No. 69.
Date of issuance: June 8, 1995.
Effective date: June 8, 1995.
Amendment No.: 78
Facility Operating License No. NPF-47. The amendment revised the
operating license.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36436) The October 18, 1993, supplemental letter provided clarifying
information and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated June 8, 1995.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: January 13, 1993, as supplemented by
letter dated June 29, 1993
Brief description of amendment: The amendment revised the River
Bend Station, Unit 1 operating license to include as a licensee,
Entergy Operations, Inc. (EOI), and to authorize EOI to use and operate
River Bend and to possess and use related licensed nuclear materials.
This amendment was originally issued on December 16, 1993 as License
Amendment No. 70.
Date of issuance: June 8, 1995
Effective date: June 8, 1995
Amendment No.: 79 [[Page 32377]]
Facility Operating License No. NPF-47. The amendment revised the
operating license.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36436) The June 29, 1993, supplemental letter provided clarifying
information and did not change the initial no significant hazards
consideration determination.The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated June 8, 1995.No
significant hazards consideration comments received. Yes. Comments and
a request for hearing were received from Cajun Electric Power
Cooperative of Baton Rouge, Louisiana.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: March 1, 1995
Brief description of amendment: The amendment revised the
surveillance criteria for certain pumps and valves in the Low Pressure
Coolant Injection (LPCI) subsystem; the Core Spray subsystems; and the
Residual Heat Removal (RHR) Service Water, High Pressure Coolant
Injection (HPCI), Emergency Service Water (ESW), and River Water Supply
systems. The surveillance criteria changed from every three months to
the testing frequency specified in the Inservice Testing program.
Date of issuance: May 18, 1995
Effective date: May 18, 1995
Amendment No.: 210
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18626) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 18, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center,
Linn County, Iowa
Date of application for amendment: March 10, 1995
Brief description of amendment: The amendment deletes Technical
Specification Sections 3.7/4.7.H.3 to eliminate redundant Limiting
Conditions of Operation and Surveillance Requirements for the
containment hydrogen and oxygen analyzers.
Date of issuance: May 31, 1995
Effective date: May 31, 1995
Amendment No.: 211
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20518) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 31, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 28, 1990
Brief description of amendment: The amendment revised the Technical
Specifications to establish periodic operability testing of the reactor
vessel overfill protection system. The changes were requested to
satisfy a commitment in the licensee's response to Generic Letter 89-
19, ``Request for Action Related to Resolution of Unresolved Safety
Issue (USI) A-47.''
Date of issuance: June 8, 1995
Effective date: June 8, 1995
Amendment No.: 169
Facility Operating License No. DPR-46. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 31, 1990 (55 FR
45885) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 8, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, NE 68305.
Northeast Nuclear Energy Company, Docket No. 50-245,
MillstoneNuclear Power Station, Unit 1, New London County,
Connecticut
Date of application for amendment: March 31, 1995
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to increase the as-found setpoint tolerance of the
safety/relief valves (SRVs) from plus or minus 1% to plus or minus 3%.
In addition, the amendment (1) allows the as-found condition of one SRV
to be inoperable, (2) clarifies the 1325 psig safety limit wording, (3)
increases the number of SRVs to be tested during each refueling outage,
(4) makes editorial changes to reflect the TS changes, and (5) revises
the bases for the applicable sections.
Date of issuance: May 31, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 82
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20520) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 31, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
MillstoneNuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: January 23, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications to modify the containment spray system by replacing the
present sodium hydroxide spray additive with the trisodium phosphate
dodecahydrate pH control agent.
Date of issuance: May 26, 1995
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 115
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11136). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 26, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: September 20, 1994, as
supplemented by letter dated April 14, 1995. [[Page 32378]]
Brief description of amendments: The proposed amendments revise
surveillance requirements (SRs) as recommended by NRC Generic Letter
(GL) 93-05, ``Line-Item Technical Specification Improvements to Reduce
Surveillance Requirements for Testing During Power Operation'' of the
combined Technical Specifications (TS) for the Diablo Canyon Nuclear
Power Plant Unit Nos. 1 and 2. The specific TS changes are as follows:
(1) TS SR 4.1.3.1.2 is revised to change the frequency for testing
the movability of the control rods from at least once per 31 days to at
least once per 92 days.
(2) TS 3/4.3.2, Table 4.3-2, ``Engineered Safety Features Actuation
System Instrumentation Surveillance Requirements,'' Functional Unit
3.c.4), and TS 3/4.3.3.1, Table 4.3-3, ``Radiation Monitoring
Instrumentation for Plant Operations SRs,'' is revised to change the
monthly channel functional test to quarterly.
(3) TS 3/4.5.1 is changed as follows: (a) TS SR 4.5.1.1a.1) is
revised to more clearly state that the accumulator water volume and
pressure must be verified to be within their limits. (b) TS SR
4.5.1.1b. is revised to specify that the boron concentration
surveillance is not required to be performed if the accumulator makeup
source was the refueling water storage tank (RWST). (c) TS SR 4.5.1.2
is relocated to plant procedures.
(4) TS SR 4.5.2c.2) is revised to clarify that a separate
containment entry to verify the absence of loose debris is not required
after each containment entry.
(5) TS SR 4.6.2.1d. is revised to change the frequency for a
containment spray header flow test from at least once per 5 years to at
least once per 10 years.
(6) TS SR 4.6.4.2a. is revised to change the verification of the
minimum hydrogen recombiner sheath temperature from at least once per 6
months to at least once each refueling interval.
(7) TS SR 4.7.1.2.1 is revised to change the surveillance frequency
for testing each auxiliary feedwater (AFW) pump from at least once per
31 days to at least once per 92 days on a staggered test basis.
(8) TS SR 4.10.1.2 is revised to lengthen the allowed period of
time for a rod drop test from 24 hours to 7 days prior to reducing
shutdown margin to less than the limits of TS 3.1.1.1.
(9) TS SR 4.11.2.6 is revised to change the surveillance frequency
from 24 hours to 7 days when radioactive material is being added to the
gas decay tanks and to add a requirement to monitor radioactive
material concentrations in the gas decay tanks at least once per 24
hours when system degassing operations are in progress.
Date of issuance: May 26, 1995
Effective date: May 26, 1995, to be implemented within 60 days of
issuance.
Amendment Nos.: Unit 1 - Amendment No. 102; Unit 2 Amendment No.
101
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53843) The April 14, 1995, letter provided clarifying information and
did not change the initial no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 26, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: December 30, 1994 (LAR 94-12)
Brief description of amendments: These amendments clarify the
technical specifications (TS) issued in license amendments 84/83
associated with the Eagle 21 reactor protection system modification,
delete TS references to RM-14A and RM-14B, remove cycle-specific TS
requirements, and incorporate editorial corrections.
Date of issuance: June 2, 1995
Effective date: June 2, 1995, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1 - Amendment No. 103; Unit 2 - Amendment No.
102
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14026) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 2, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: February 6, 1995, as
supplemented by letters dated March 23, 1995, and May 22, 1995.
Brief description of amendments: The amendments would allow the
storage of fuel with enrichments up to and including 5.0 weight percent
U-235, would clarify that substitution of fuel rods with filler rods is
acceptable for fuel designs that have been analyzed with applicable
NRC-approved codes and methods, and would allow the use of ZIRLO fuel
cladding in the future in addition to Zircaloy-4.
Date of issuance: June 7, 1995
Effective date: June 7, 1995, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1 - Amendment No. 104; Unit 2 - Amendment No.
103
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11138) The licensee's supplemental letters provided additional
clarifying information. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated June 7, 1995. No
significant hazards consideration comments received: Yes. Comments were
submitted by Jill ZamEk on behalf of the San Luis Obispo Mothers for
Peace by letter dated March 30, 1995.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay
Power Plant, Unit 3, Humboldt County, California
Date of application for amendment: November 23, 1994, as
supplemented April 27, 1995.
Brief description of amendment: This amendment revised the
Technical Specifications Section VII.C., Plant Staff, to decrease the
minimum staff requirements for the shift operating organization from
five to two persons.
Date of issuance: May 31, 1995
Effective date: This license amendment is effective as of the date
of [[Page 32379]] its issuance and must be fully implemented no later
than 30 days from the date of issuance.
Amendment No.: 28Facility License No. DPR-7: The amendment revised
the TS.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11139) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 31, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501.
PECO Energy Company, Public Service Electric and Gas
CompanyDelmarva Power and Light Company, and Atlantic City Electric
Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: February 10, 1995
Brief description of amendments: These amendments correct
administrative errors in Section 4.11.A of the Technical Specifications
(TSs). The errors were made in the TSs by Amendments 9 and 7 dated June
25, 1975.
Date of issuance: May 30, 1995
Effective date: May 30, 1995
Amendments Nos.: 202 and 205
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20521) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 30, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas
CompanyDelmarva Power and Light Company, and Atlantic City Electric
Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station,Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: September 26, 1994
Brief description of amendments: These amendments extend the
surveillance test intervals and allowable out-of service times for the
testing and or repair of instrumentation that actuate the Reactor
Protection System, Primary Containment Isolation, Core and Containment
Cooling systems, Control Rod Blocks, Radiation Monitoring systems and
Alternate Rod Insertion/Recirculation Pump Trip.
Date of issuance: June 6, 1995
Effective date: June 6, 1995
Amendments Nos.: 203 and 206
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14027) The supplemental letters dated January 5, and March 23, 1995,
provided clarifying information and did not change the initial proposed
no significant hazards consideration determination.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated June 6, 1995.No significant hazards consideration
comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: April 30, 1993
Brief description of amendments: These amendments changed the
Technical Specifications by deleting Section 3/4.3.8 of the Turbine
Overspeed Protection System.
Date of issuance: June 1, 1995
Effective date: June 1, 1995
Amendment Nos.: 146 and 116
Facility Operating License Nos. NPF-14 and NPF-22. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 9, 1993 (58 FR
32389) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 1, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: August 25, 1993, as
supplemented by letters dated June 27, 1994, and May 5, 1995
Brief description of amendments: These amendments modify Technical
Specification Surveillance Requirement 4.7.1.3 to require that all
spray pond spray network piping above the frost line be drained at an
ambient temperature below 40 deg.F, and within 1 hour after being used
only when the ambient air temperature is below 40 deg.F.
Date of issuance: June 1, 1995
Effective date: June 1, 1995
Amendment Nos.: 90 and 54
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50972) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 1, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: August 30, 1994
Brief description of amendment: The changes relocate Technical
Specification (TS) 3.3.7.9, Loose Parts Detection System (LPDS),
Surveillance Requirement 4.3.7.9, and associated Bases from the TSs to
the Updated Final Safety Analysis Report. The TS index is also revised
by removing the reference to LPDS.
Date of issuance: May 25, 1995
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 73
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16197) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 25, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: February 9, 1995
Brief description of amendments: The amendments revise the
Administrative [[Page 32380]] Controls section of the Technical
Specifications to reflect organizational changes and resultant
management title changes.
Date of issuance: June 6, 1995
Effective date: June 6, 1995
Amendment Nos.: 168 and 150
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16200) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 6, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama.
Date of application for amendments: March 6, 1995
Brief description of amendments: The amendments relocate the
seismic and meteorological monitoring instrumentation from the
Technical Specifications to the Final Safety Analysis Report in
accordance with the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors,'' dated July
22, 1993.
Date of issuance: May 22, 1995
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 115 and 107
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18628) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 22, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear
Plant, Unit 1, Hamilton County, Tennessee
Date of application for amendment: April 6, 1995 (TS 95-09)
Brief description of amendment: The amendment modifies Operating
License Condition 2.C.(25) to provide a limited extension of the ice
condenser surveillance test interval on Unit 1 to coincide with the
Cycle 7 refueling outage.
Date of issuance: May 30, 1995
Effective date: May 30, 1995
Amendment No.: 200
Facility Operating License Nos. DPR-77: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20526) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 30, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 6, 1995
Brief description of amendments: The amendments revise the
surveillance requirement for the power range neutron flux channel
calibration frequency from monthly to every 31 effective full power
days and delays first performance of the surveillance after reaching 15
percent power for 96 hrs.
Date of issuance: May 30, 1995
Effective date: May 30, 1995
Amendment Nos.: 199 and 190
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20530) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 30, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: April 6, 1995
Brief description of amendments: The amendments revise the
definition of core alteration, quadrant power tilt ratio, and modifies
the operational mode parameters table in the Unit 1 technical
specifications.
Date of issuance: June 1, 1990
Effective date: June 1, 1990
Amendment Nos.: 201 and 191
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20531) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 1, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: February 14, 1994 (TXX-94046 LAR 94-006)
Brief description of amendments: The proposed changes would revise
the Technical Specifications (TSs) for Comanche Peak Steam Electric
Station, Units 1 and 2 in the following three areas: 1) a change to the
allowable value for the Unit 2 pressurizer pressure-low and Unit 2
overtemperature N-16 (OTN-16) reactor trip setpoints; 2) an
administrative change to delete an option which allowed continued
operation for a period of time when a reactor trip system (RTS) or
engineered safety features actuation system (ESFAS) instrumentation or
interlocks trip setpoint is found less conservative than the allowable
value; and 3) an administrative change to combine the Unit 1 and Unit 2
line items for RTS or ESFAS trip setpoint and allowable values which
are the same.
Date of issuance: May 31, 1995
Effective date: May 31, 1995, to be implemented within 30 days.
Amendment Nos.: Unit 1 - Amendment No. 41; Unit 2 - Amendment No.
27
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32238) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 31, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: March 24, 1995
Brief description of amendment: The amendment relaxes the
requirement to [[Page 32381]] sample the accumulator after refilling
from the RWST.
Date of issuance: May 30, 1995
Effective date: May 30, 1995, to be implemented within 30 days of
issuance.
Amendment No.: Amendment No. 87
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18632) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 30, 1995.No significant
hazards consideration comments received: No. Local Public Document Room
locations: Emporia State University, William Allen White Library, 1200
Commercial Street, Emporia, Kansas 66801 and Washburn University School
of Law Library, Topeka, Kansas 66621.
Dated at Rockville, Maryland, this 14th day of June, 1995.
For the Nuclear Regulatory Commission
John N. Hannon,
Acting Deputy Director, Division of Reactor Projects - III/IV, Office
of Nuclear Reactor Regulation
[Doc. 95-15057 Filed 6-20-95; 8:45]
BILLING CODE 7590-01-F
This is a copy of a public record, reproduced as it was published. It is not legal advice, and it may not be the version a court would rely on. Check the official source before you cite it.