[Federal Register Volume 62, Number 175 (Wednesday, September 10, 1997)]
[Notices]
[Pages 47696-47705]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-23820]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Pub. L. 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
[[Page 47697]]
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 August 18, 1997, through August 28, 1997.
The last biweekly notice was published on August 27, 1997 (62 FR
45452).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By October 10, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any
[[Page 47698]]
hearing held would take place before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: August 15, 1997.
Description of amendment request: The proposed amendment would
revise portions of the facility Technical Specifications regarding
facility staffing and training requirements to power operations. By
letter dated August 7, 1997, the licensee certified permanent cessation
of power operations and permanent removal of fuel from the reactor
vessel. By two letters both dated August 15, 1997, the licensee has
also submitted a related ``Request for Exemption from Certain
Requirements of 10 CFR 50.54, Conditions of License,'' and a ``Request
for Approval of the Certified Fuel Handler Training and Retraining
Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
The proposed change does not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The purpose of the proposed change is to eliminate the
requirements for licensed operators and a licensed operator training
program and to replace those with certified fuel handlers and a
certified fuel handler training and retraining program. Since the
plant has permanently ceased operation and will be maintained in a
defueled condition, the range of accidents for which an operator
needs to be trained has significantly diminished such that a
training program of the depth and breadth of that required by 10 CFR
[Part] 55 is no longer needed. In lieu of a 10 CFR [Part] 55
licensed operator training program, a[n] NRC-approved certified fuel
handler training and retraining program will be implemented. Since
this training program will adequately equip appropriate operations
personnel for fuel handling operations, including responses to
abnormal events/accidents, there will be no increase in the
probability of these events occurring or in the consequences of
these events. The proposed changes do not affect plant equipment or
the procedures for equipment operation or response to abnormal
events/accidents.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The purpose of this proposed change is to eliminate the
requirements for licensed operators and a licensed operator training
program and to replace those with certified fuel handlers and a
certified fuel handler training and retraining program. This change
ensures the qualifications of operations personnel are commensurate
with the tasks to be performed and the conditions to be responded
to. This change does not affect plant equipment or the procedures
for operating plant equipment and, therefore, does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change is to eliminate the requirements for
licensed operators and a licensed operator training program to
replace those with certified fuel handlers and a certified fuel
handler training and retraining program. This change ensures the
qualifications of the operations personnel are commensurate with the
tasks to be performed and the conditions to be responded to. The
assumptions for a fuel handling accident in the Fuel Building are
not affected by the proposed changes. Therefore, the proposed
amendment does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011.
NRC Acting Project Director: Ronald B. Eaton.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: August 1, 1997.
Description of amendment request: The amendments would change
Technical Specification Section 4.2.1 of Appendix B to the licenses.
The changes include rewording of the section to generically state that
Public Service Gas & Electric (PSE&G) will adhere to the Section 7,
Incidental Take Statement, approved by the National Marine Fisheries
Service (NMFS). Removing the specific requirements of this section
enables PSE&G to utilize relief granted by the NMFS on a case-by-case
basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes are administrative in nature and would in no way
affect the initial conditions, assumptions, or conclusions of the
Salem [Nuclear] Generating Station, Units 1 and 2, accident
analyses. In addition, the proposed changes would not affect the
operation or performance of any equipment assumed in the accident
analyses. Based on the above information, we conclude that the
proposed changes would not significantly increase 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 changes are administrative in nature and would in no way
impact or alter the configuration or operation of the facilities and
would create no new modes of operation. We therefore conclude that
the proposed changes would not create the possibility of a new or
different kind of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
As indicated in the discussion of Criterion 1, the changes are
administrative in nature and would in no way affect plant or
equipment operation or the accident analysis. We therefore conclude
that the proposed changes would not result in 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
[[Page 47699]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendments request: July 23, 1997.
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TSs) by relocating the reactor
coolant system pressure and temperature limits from the TSs to the
proposed Pressure Temperature Limits Report in accordance with the
guidance provided by Generic Letter 96-03, ``Relocation of the Pressure
Temperature Limit Curves and Low Temperature Overpressure Protection
System 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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed removal of the Reactor Coolant System (RCS)
pressure temperature (P/T) limits from the Technical Specifications
(TSs) and relocation to the proposed Pressure Temperature Limits
Report (PTLR) in accordance with the guidance provided by Generic
Letter (GL) 96-03 is administrative in that the requirements for the
P/T limits are unchanged. The P/T limits proposed for inclusion in
the PTLR are based on the fluence associated with 2775 MW
[megawatts] thermal power and operation through 36 effective full
power years (EFPY). GL 96-03 requires that the P/T limits be
generated in accordance with the requirements of 10 CFR [Part] 50,
Appendices G and H, documented in an NRC-approved topical report
incorporated by reference in the TSs. Accordingly, the proposed
curves have been generated using the NRC-approved methods described
in WCAP-14040-NP-A, Revision 2, and meet the requirements of 10 CFR
[Part] 50, Appendices G and H. TS 3.4.10.1 will continue to require
that the RCS pressure and temperature be limited in accordance with
the limits specified in the PTLR. The NRC-approved methodology for
generating the P/T limit, WCAP-14040-NP-A, Revision 2, will be
specified in TS 6.9.1.15 and NRC approval will be required in the
form of a TS Amendment prior to changing the methodology. Use of P/T
limit curves generated using the NRC-approved methods described in
WCAP-14040-NP-A, Revision 2, as specified in TS 6.9.1.15, will
provide additional protection for the integrity of the reactor
vessel, thereby assuring that the reactor vessel is capable of
providing its function as a radiological barrier.
TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit 1 and Unit 2
provides the operability requirements for RCS low temperature
overpressure protection (LTOP). Specifically, TS 3.4.10.3 requires
that two residual heat removal (RHR) system suction relief valves
(RHRRVs) be operable or that the RCS be vented at RCS cold leg
temperatures less than or equal to 310 deg.F. GL 96-03 recognizes
that RHRRVs do not have variable pressure lift setpoints and states
that those plants that rely on the RHRRVs for LTOP should continue
to address the LTOP requirements in the TS. Consistent with GL 96-
03, the Farley Unit 1 and Unit 2 requirements for LTOP will be
retained in TS 3.4.10.3.
Based on the above evaluation, the proposed changes are
administrative in nature and do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As stated above, the proposed changes to remove the RCS P/T
limits from the TSs and relocate them to the proposed PTLR is an
administrative change. Consistent with the guidance provided by GL
96-03, the proposed P/T limits contained in the proposed PTLR meet
the requirements of 10 CFR [Part] 50, Appendices G and H, and were
generated using the NRC-approved methods described in WCAP-14040-NP-
A, Revision 2. The proposed changes do not result in a physical
change to the plant or add any new or different operating
requirements on plant systems, structures, or components with the
exception of limiting the number of operating RCPs [reactor coolant
pumps] at RCS temperatures below 110 deg.F. Limiting the number of
operating RCPs below 110 deg.F results in a reduction in the
[delta]P between the reactor vessel beltline and the RHRRVs, thereby
providing additional margin to limits of Appendix G. Provisions are
made to allow the start of a second RCP at temperatures below
110 deg.F in order to secure the pump that was originally operating
without interrupting RCS flow. The LTOP enable temperature exceeds
the minimum LTOP enable temperature determined using the NRC-
approved methods described in WCAP-14040-NP-A, Rev. 2, thereby
providing additional assurance that the LTOP system will be
available to protect the RCS in the event of an overpressure
transient at RCS temperatures at or below 310 deg.F. Using the
methods contained in WCAP-14040-NP-A, Rev. 2, the minimum boltup
temperature for the reactor vessel flange region is 60 deg.F which
is less than the design limits of the fuel cladding. Administrative
controls require a minimum RCS temperature of 68 deg.F when fuel is
loaded in the reactor vessel to protect against brittle failure of
the fuel cladding, and also require that the component cooling water
(CCW) temperature be maintained between 60 deg.F and 105 deg.F
during refueling operations, thus reducing the potential for the RCS
temperature to be less than the minimum boltup temperature specified
in the proposed PTLRs.
As stated in the above response, implementation of the proposed
changes do not result in a significant increase in the probability
of a new or different accident (i.e., loss of reactor vessel
integrity). The RCS P/T limits will continue to meet the
requirements of 10 CFR [Part] 50, Appendices G and H, and will be
generated in accordance with the NRC approved methodology described
in WCAP-14040-NP-A, Rev. 2. Therefore, the proposed changes do not
result in a significant increase in the possibility of a new or
different accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is not affected by the removal of the RCS
P/T limits from the TSs and relocating them to the proposed PTLR.
The RCS P/T limits will continue to meet the requirements of 10 CFR
[Part] 50, Appendices G and H. To provide additional assurance that
the P/T limits continue to meet the requirements of Appendices G and
H, TS 6.9.1.15 will require the use of the NRC-approved methodology
described in WCAP-14040-NP-A, Rev. 2, to generate P/T limits. The
RCS LTOP requirements will be retained in TS 3.4.10.3 due to use of
the RHRRVs for LTOP, consistent with the guidance provided by GL 96-
03. The LTOP enable temperature exceeds the LTOP enable temperature
determined in accordance with the NRC-approved methodology, thus
protecting the RCS in the event of a low temperature overpressure
transient over a broader range of temperatures than required by
WCAP-14040-NP-A, Rev. 2. Administrative procedures preclude
operation of the RCS at temperatures below the minimum boltup
temperature for the reactor vessel head, thus precluding the
possibility of tensioning the reactor vessel head at RCS
temperatures below the minimum boltup temperature. Operation of the
plant in accordance with the RCS P/T limits specified in the PTLR
and continued operation of the LTOP system in accordance with TS
3.4.10.3 will continue to meet the requirements of 10 CFR [Part] 50,
Appendices G and H, and will therefore, assure that a margin of
safety is not significantly decreased as the result of the proposed
changes.
Based on the preceding analysis, SNC [Southern Nuclear Operating
Company, Inc.] has determined that removal of the RCS P/T limits
from the TS and relocation to the proposed PTLR will not
significantly increase the probability or consequences of an
accident previously evaluated, create the possibility of a new or
different kind of accident from any accident previously evaluated,
or involve a significant reduction in a margin of safety. SNC
therefore concludes that the proposed change meets the requirements
of 10 CFR 50.92(c) and does
[[Page 47700]]
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: Houston-Love Memorial Library, 212 W.
Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302. Attorney
for licensee: M. Stanford Blanton, Esq., Balch and Bingham, Post Office
Box 306, 1710 Sixth Avenue North, Birmingham, Alabama 35201. NRC
Project Director: Herbert N. Berkow.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 20, 1997 (TS-97-004).
Description of amendment request: The proposed amendment would be
an administrative change that would revise the analytical methodology
used to determine the low temperature overpressure protection (LTOP)
event heatup and cooldown curves. This revised methodology would be
incorporated by reference in the Watts Bar Nuclear Plant (WBN), Unit 1
Technical Specification (TS) 5.9, ``Reporting Requirements,'' Section
5.9.6, ``Reactor Coolant System (RCS) Pressure and Temperature Limits
Report (PTLR),'' upon approval for use by the U.S. Nuclear Regulatory
Commission (NRC). The revised methodology extends the current LTOP
requirements through the end of 7 effective full power years (EFPY).
The only technical change being proposed is the substitution of the 7
EFPY American Society of Mechanical Engineering (ASME), Appendix G,
heatup and cooldown curves adjusted by ASME Code Case N-514, ``Low
Temperature Overpressure Protection'' in place of the current 1.5 EFPY
curves as the bounding curves for the LTOP setpoints. This change will
not impact the current 10 CFR 50, Appendix G, pressure/temperature (P/
T) limit curves used for heatup and cooldown that are based on 7 EFPY.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The Nuclear Regulatory Commission has provided standards for
determining whether a significant hazards consideration exists (10
CFR 50.92). A proposed amendment to an operating license for a
facility 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 below for the proposed
amendment.
(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 LTOP setpoints (identified as the cold overpressure
mitigation system (COMS) for WBN), adjusted for instrument
inaccuracy, pressure differential, and setpoint overshoot by the
scaling and setpoint documents (SSDs), ensure that the 10 CFR 50,
Appendix G P/T [pressure and temperature] limits based on 7 EFPY are
not exceeded by more than the provisions of ASME Code Case N-514,
and therefore, ensure that the RCS integrity is maintained.
The change does not modify the RCS pressure boundary, nor make
any physical changes to the facility design, material, construction
standards, or setpoints. The LTOP enabling temperature based on TS
3.4.12, ``Cold Overpressure Mitigation System (COMS),'' is [less
than or equal to] 350 degrees F and is more conservative than a
value of 271.1 degrees F (RTNDT + 90 degrees F) based on
7 EFPY. This temperature would be acceptable based on NRC Branch
Technical Position-Reactor Systems Branch (BTP-RSB)-5.2,
``Overpressurization Protection of Pressurized Water Reactors While
Operating at Low Temperatures.'' The LTOP enabling temperature
remains unchanged by this proposed amendment. The probability of a
LTOP event occurring is independent of the P/T limits for the RCS
pressure boundary; therefore, the probability of an LTOP event
occurring remains unchanged.
The calculation of the P/T limits in accordance with approved
regulatory methods based on 7 EFPY provides assurance that reactor
pressure vessel fracture toughness requirements are met and the
integrity of the RCS pressure boundary is maintained. LTOP setpoints
based on 1.5 EFPY P/T limits have provided margin such that a
pressure excursion exceeding the 7 EFPY limits would not exceed the
1.5 EFPY limits. This margin between the 7 EFPY curves and the LTOP
setpoints is maintained by changing the bounding curves for the LTOP
setpoints to 7 EFPY curves adjusted by the provisions of ASME Code
Case N-514. The only technical change being made is the bounding
curves which provide the basis for the current LTOP setpoints.
The use of theoretical fluence for generating the P/T curves to
be used for the first 7 EFPY is appropriate and was submitted July
31, 1995, with the WBN Unit 1 PTLR, Revision 4 and WCAP-13829,
Revision 2, ``Heatup and Cooldown Limit Curves for Normal Operation
for Watts Bar Unit 1.'' The present 7 EFPY curves are generated
using a theoretical value for fluence calculated by Westinghouse in
accordance with NRC approved methodology since WBN had no
surveillance capsule data available at the time of plant startup.
This value for fluence is conservative, and the actual fluence to
the intermediate shell forging (the controlling beltline material)
is expected to be significantly less than the theoretical value used
to generate the initial 7 EFPY curves since WBN is transitioning to
a low-leakage core. The LTOP bounding curves are based on 7 EFPY
curves adjusted in accordance with ASME Code Case N-514 which were
generated using the same theoretical fluence as used for the P/T
curves. The significance of using the theoretical value of fluence
in generating these curves is the additional margin that exists
between the 7 EFPY theoretical curves and curves that would be
generated using actual fluence values from capsule data. This
additional margin reduces the significance of changing the LTOP
basis from the 1.5 EFPY curves to the 7 EFPY curves adjusted for
ASME Code Case N-514.
This change does not adversely affect the integrity of the RCS
such that its function in the control of radiological consequences
is affected. In addition, the change does not affect any fission
barrier. The change does not degrade or prevent the LTOP power
operated relief valves (PORVs) or other safety related systems from
responding to accidents described in Chapter 15 of the Final Safety
Analysis Report (FSAR). In addition, the change does not alter any
assumptions previously made in the radiological consequences of an
accident described in the FSAR. Therefore, the consequences of an
accident previously evaluated in the FSAR are not increased. Thus,
the operation of WBN Unit 1 in accordance with this proposed
amendment does not involve a significant increase in the probability
or consequences of any 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 Appendix G P/T limitations were prepared using methods
derived from the ASME Boiler and Pressure Vessel Code Section III
and the criteria set forth in NRC Regulatory Standard Review Plan
5.3.2, ``Pressure-Temperature Limits.'' The use of ASME Code Case N-
514 and the theoretical fluence value for 7 EFPY does not modify the
RCS pressure boundary, nor make any physical changes to the LTOP
setpoints or system design. The proposed change was prepared in
accordance with regulatory requirements and provides evaluation of
LTOP events based on 7 EFPY theoretical fluence which is more
limiting than actual expected neutron exposure for that same period.
This proposed change is an administrative change which
incorporates by reference the use of an NRC approved methodology;
therefore, the change does not cause the initiation of any accident
nor create any new creditable limiting failure for safety-related
[[Page 47701]]
systems and components. The change does not result in an event
previously deemed incredible being made credible. As such, it does
not create the possibility of an accident different than any
evaluated in the FSAR.
The change does not have any effect on the ability of the
safety-related systems to perform their intended safety functions.
The change does not create failure modes that could adversely impact
safety-related equipment. Therefore, it will not create the
possibility of a malfunction of equipment important to safety
different than previously evaluated in the FSAR. Thus, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different type 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 margin of
safety.
The 10 CFR 50, Appendix G P/T limitations were prepared using
methods derived from ASME Section III and criteria set forth in NRC
Regulatory Standard Review Plan 5.3.2. These documents along with
the calculational limitations specified in 10 CFR 50.61 are an
acceptable method for implementing the requirements of 10 CFR 50
Appendices G and H. Inherent conservatisms in the P/T limits
resulting from these documents include:
a. An assumed defect in the reactor vessel wall with a depth
equal to \1/4\ of the thickness (T) of the vessel wall and a length
equal to 1\1/2\ times the thickness of the vessel wall.
b. Assumed reference flaw oriented in both longitudinal and
circumferential directions and limiting material property. At WBN,
the only weld in the core region is oriented in the circumferential
direction.
c. A factor of safety of 2 is applied to the membrane stress
intensity factor.
d. The limiting toughness is based upon a reference value
(KIM) which is the lower bound of the dynamic crack
initiation and arrest toughness.
e. A 2-sigma margin term is applied in determining the adjusted
reference temperature (ART) that is used in calculating the limiting
toughness.
Beyond the conservatisms described above, WBN has the following
additional margin:
a. The value of fluence used in the calculation of the WBN Unit
1 Appendix G P/T limits is a theoretical value calculated by NRC
approved methodology.
b. The ART for 7 EFPY is based on the theoretical value for
fluence and therefore is conservative. The LTOP enabling temperature
of [less than or equal to] 350 degrees F in accordance with TS
3.4.12 is conservative with respect to (RTNDT + 90
degrees F) which based on an ART of 181.1 degrees F would equal
271.1 degrees F. An enabling temperature of (RTNDT + 90
degrees F) is based on NRC BTP-RSB 5.2.
The ASME Working Group for Operating Plant Criteria developed
Code Case N-514 as an alternative methodology to the safety margin
requirements of Appendix G to 10 CFR 50. The Code Case provides
criteria to determine pressure limits during LTOP events that avoid
certain operational restrictions, provide adequate margins against
failure of the reactor vessel, and reduce the potential for
unnecessary activation of the relief valves used for LTOP.
Specifically, the N-514 Code Case allows determination of the LTOP
setpoints such that for LTOP events the maximum pressure in the
reactor vessel would not exceed 110% of the P/T limits of the
existing ASME Appendix G curves, and redefines the enabling
temperature as a coolant temperature less than 200 degrees F or a
reactor vessel metal temperature less than RTNDT + 50
degrees F. Code Case N-514 has been approved by the ASME Code
Committee and its content has been incorporated in Appendix G of
ASME Section XI and published in the 1993 Addenda and 1995 Edition.
Code Case N-514 has not been approved for use in Regulatory Guide
1.147, ``Inservice Inspection Code Case Acceptability, ASME Section
XI;'' however, it has been included in the Draft Regulatory Guide
1.147 (Task DG-1050) which is currently out for public review and
comment. As stated above, WBN Unit 1 uses Appendix G for the P/T
limits for plant operation and an LTOP enabling temperature greater
than RTNDT + 90 degrees F which is more conservative than
the alternative methodology contained in Code Case N-514.
The need for implementation of Code Case N-514 at WBN involves
the avoidance of certain operational restrictions associated with
low temperature operation of the plant. Use of Appendix G P/T limits
to determine the PORV setpoints would result in pressure setpoints
within the operating window; consequently, no margin would be
available for normal operating pressure surges. Therefore, operating
with these limits could result an unnecessary challenge to the PORVs
and cavitation of the reactor coolant pumps (RCP) during normal
operation. Additionally, the need to raise the RCS inventory by
external heating methods to a temperature high enough to avoid PORV
activation when starting a RCP from a RCS cold shutdown condition
could result in undesirable thermal transients in the RCS.
Utilizing the methodology set forth in the ASME Boiler and
Pressure Vessel Code Section XI, Appendix G, which includes the
provisions of Code Case N-514, NRC Regulatory Standard Review Plan
5.3.2, 10 CFR 50.61, and 10 CFR 50, Appendices G and H with the
above additional margins ensures that proper limits and conservative
safety factors are maintained. Thus the proposed change does not
significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
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.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: March 18, 1997.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to increase the High Pressure Coolant
Injection (HPCI)
[[Page 47702]]
system low pressure isolation setpoint from greater than 80 psig to
greater than 100 psig.
Date of issuance: August 21, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 161, 156.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17228).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: June 12, 1997.
Brief description of amendments: The amendments change the name
``Duke Power Company'' to ``Duke Energy Corporation'' in the Catawba
operating licenses and appendices as a result of Duke Power Company's
recent name change.
Date of issuance: August 22, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 161 and 153.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35848).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 22, 1997, and an Environmental
Assessment dated July 31, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 12, 1997.
Brief description of amendments: The amendments change the name
``Duke Power Company'' to ``Duke Energy Corporation'' in the McGuire
operating licenses and appendices as a result of Duke Power Company's
recent name change.
Date of issuance: August 26, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 176 and 158.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Licenses.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35848).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 26, 1997. An Environmental
Assessment was issued and dated August 15, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of amendment request: October 26, 1995 and supplemented by
letters dated April 7 and July 30, 1997.
Brief description of amendment: The amendment revised the technical
specifications for 16 editorial changes and deletes the reuirement for
a program to prevent and detect Asiatic Clams (Corbicula) in the
service water system (SWS). The Corbicula program is no longer needed
because the facility has been modified and SWS no longer takes water
from the Mississippi River; source of the larvae and infestation.
Date of issuance: August 26, 1997.
Effective date: August 26, 1997.
Amendment No.: 95.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62492).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 26, 1997.
No significant hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of amendment request: November 15, 1996, as supplemented May 9
and August 15, 1997.
Brief description of amendment: The amendment revises the technical
specifications to increase the two recirculation loop Minimum Critical
Power Ratio (MCPR) from 1.07 to 1.10 and the single recirculation loop
MCPR limit from 1.08 to 1.12.
Date of issuance: August 26, 1997.
Effective date: August 26, 1997.
Amendment No.: 96.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
127).
The May 9 and August 15, 1997, submittal provided clarifying
information that did not change the initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 26, 1997.
No significant hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of amendment request: January 20, 1997 as supplemented by
letter dated July 7, 1997.
Brief description of amendment: The amendment revises the technical
specifications to allow the use of flow control spectral shift
strategies to increase cycle energy. The revision is based on a Maximum
Extended Load Line Limit (MELLL) analysis for the River Bend Station.
Date of issuance: August 26, 1997.
Effective date: August 26, 1997.
Amendment No.: 97.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications/operating license.
Date of initial notice in Federal Register: February 26, 1997 (62
CFR 8799).
The July 7, 1997 submittal provided clarifying information and did
not change the initial no significant hazards consideration
determination.
[[Page 47703]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of amendment request: November 6, 1996, as supplemented by
letter dated July 31, 1997.
Brief description of amendment: The amendment revises the Technical
Specifications to delete the requirement for the Penetration Valve
Leakage Control System. The licensee requested deferal of the proposal
to increase the allowed leakage by main steam isolation valves and to
delete the requirement for the Main Steam Positive Leakage Control
System.
Date of issuance: August 26, 1997.
Effective date: August 26, 1997.
Amendment No.: 98.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
125).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power & Light
Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1,
Claiborne County, Mississippi
Date of application for amendment: October 22, 1996, as
supplemented by letter dated June 26, 1997.
Brief description of amendment: The amendment revises Figure
3.4.11-1, ``Minimum Reactor Vessel Metal Temperature vs. Reactor Vessel
Pressure,'' in Limiting Condition for Operation 3.4.11, ``RCS [Reactor
Coolant System] Pressure and Temperature (P/T) Limits,'' of the
Technical Specifications. The previous figure was only up to 10
Effective Full Power Years (EFPYs) and this amendment revises the
figure up to 32 EFPYs. There are now five curves of Figure 3.4.11-1 for
five different EFPY periods: up to 16, 16 to 20, 20 to 24, 24 to 28,
and 28 to 32. The licensee submitted two sets of curves. The first set
replaced TS Figure 3.4.11-1. The second set were duplicates of the
first set except the second set also contained detailed information
used in development of the curves and would be included in the next
update of the Updated Final Safety Analysis Report. There were also
minor additions to Surveillance Requirements (SRs) 3.4.11.1 and
3.4.11.2 to have the SRs reference the ``applicable Figure 3.4.11-1
based on the current effective full power year (EFPY).''
Date of issuance: August 27, 1997.
Effective date: August 27, 1997.
Amendment No: 132.
Facility Operating License No. NPF-29: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 26, 1997 (62
FR 8797).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 27, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: April 11, 1997.
Brief description of amendment: The amendment modifies Technical
Specifications 3.3.3.7.3, and Surveillance Requirements (SR) 4.3.3.7.3
for the broad range gas detection system. Also it includes some changes
to the Bases in Section 3/4.3.3.7 to incorporate information associated
with the proposed modifications. The licensee is planning to replace
the existing toxic gas monitors in the system with a new, more advanced
gas monitors.
Date of issuance: August 19, 1997.
Effective date: August 19, 1997, to be implemented within 90 days.
Amendment No.: 133.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24987)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin
I. Hatch Nuclear Plant, Unit 1, Appling County, Georgia
Date of application for amendment: April 29, 1997, as supplemented
by letter dated May 28, 1997.
Brief description of amendment: The amendment revises Hatch Unit 1
reactor vessel pressure and temperature limits to reflect data
collected from the material sample recovered during the March 1996 Unit
1 outage.
Date of issuance: August 19, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 207.
Facility Operating License No. DPR-57: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38138).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: October 4, 1996, as supplemented
June 10 and August 15, 1997 (TSCR 250).
Brief description of amendment: The amendment changes the Safety
Limit Minimum Critical Power Ratio and as a result, the operating
Minimum Critical Power Ratio. The amendment also capitalized certain
definitions and provided a uniform type font for Sections 2.1 and 3.10.
Date of Issuance: August 26, 1997.
Effective date: August 26, 1997, with full implementation within 30
days.
Amendment No.: 192.
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57484).
The Commission's related evaluation of this amendment is contained
in a
[[Page 47704]]
Safety Evaluation dated August 26, 1997.
The June 10 and August 15, 1997, submittals provided clarifying
information that did not alter the staff's initial proposed no
significant hazards considerations determination.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: April 21, 1997, as supplemented
July 17, 1997.
Brief description of amendment: The amendment reduces the required
volume of borated water in each core flood tank from 1040 ft \3\ to 940
ft \3\, reduces the required high pressure injection pump flowrate from
500 gallons per minute (gpm) to 431 gpm, and deletes the local manual
valve operability option for decay heat system valves DH-V-6A and DH-V-
6B.
Date of issuance: August 27, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 203.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 21, 1997 (62 FR
27795).
The July 17, 1997, submittal provided clarifying information that
did not alter the initial no significant hazards determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated August 27, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: April 22, 1997.
Brief description of amendments: The proposed amendment revised
Technical Specifications 5.3.1, Fuel Assemblies, and 6.9.1.6, Core
Operating Limits Report, to allow use of an alternate zirconium-based
fuel cladding, ZIRLO, and limited substitution of fuel rods by ZIRLO
filler rods.
Date of issuance: August 19, 1997.
Effective date: August 19, 1997.
Amendment Nos.: Unit 1--Amendment No. 89; Unit 2--Amendment No. 76.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 21, 1997 (62 FR
27795).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: November 20, 1995.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by providing clarifications to the applicability
and action statements in TS Table 3.3-12 relating to the Steam
Generator Blowdown Monitor and the Condensate Polishing Facility Waste
Neutralizing Sump radiation monitor.
Date of issuance: August 26, 1997.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 207.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65683).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: May 1, 1997.
Brief description of amendment: Technical Specifications 3/4.8.2.2
and 3/4.8.3.2 specify which electrical power systems are required to be
operable in Modes 5 and 6. The amendment clarifies the requirements by
identifying the specific equipment required and their alignments in
Modes 5 and 6.
Date of issuance: August 21, 1997.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 146.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30637).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: May 5, 1997.
Brief description of amendment: Technical Specification
Surveillance 4.5.2.b.1 requires that the emergency core cooling system
piping be verified full of water at least once per 31 days. The
amendment revises the surveillance to exempt the operating charging
pump(s) and associated piping from the requirement to be verified full
of water and moves the description of the verification method from the
surveillance to the Bases section.
Date of issuance: August 28, 1997.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 147.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30638).
The Commission's related evaluation of the amendment is contained
in a
[[Page 47705]]
Safety Evaluation dated August 28, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope.
PECO Energy Company, Public Service Electric and Gas Company Delmarva
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: March 31, 1997, as supplemented
by letter dated June 25, 1997.
Brief description of amendments: These amendments extend the APRM
flow bias instrumentation surveillance interval from 18 months to 24
months. This will eliminate the need to perform on-line APRM
surveillance testing, which requires plant operators to place an
operating unit in a half scram configuration.
Date of issuance: August 19, 1997.
Effective date: Units 2 and 3 effective as of date of issuance.
Amendments Nos.: 219 and 222.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 7, 1997 (62 FR
24988).
The supplemental letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 19, 1997.
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,
PA 17105.
Southern Nuclear Power Company, Inc., 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: June 13, 1997, as supplemented
by letter dated July 18, 1997.
Brief description of amendments: The amendments revise the
pressurizer safety relief valve setpoint specified in Technical
Specification 3.4.10.
Date of issuance: August 26, 1997.
Effective date: As of the date of issuance to be implemented for
Unit 1 prior to or after initial entry into Mode 3 (in accordance with
the provisions of the note to the Applicability for LCO 3.4.10)
following the fall 1997 refueling outage; for Unit 2 prior to or after
initial entry into Mode 3 (in accordance with the provisions of the
note to the Applicability for LCO 3.4.10) following the spring 1998
refueling outage.
Amendment Nos.: 98 and 76.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38139).
The supplemental material did not change the no significant hazards
finding or expand the scope of the Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830.
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: December 7, 1994 (TXX-94326), as
supplemented by letter dated June 21, 1996 (TXX-96384).
Brief description of amendments: These changes revised Section
3.7.1.5 of the Technical Specification to increase the Allowed Outage
Time for one inoperable Main Steam Isolation Valve (MSIV) while in Mode
1, and to clarify requirements related to inoperable MSIVs while in
Modes 2 and 3.
Date of issuance: August 18, 1997.
Effective date: August 18, 1997, to be implemented within 60 days.
Amendment Nos.: 54 and 40.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6312).
The additional information contained in the supplemental letter
dated June 21, 1996, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 18, 1997.
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.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: April 14, 1997 (TSCR 197), as
supplemented on August 11, 1997.
Brief description of amendments: These amendments revise Technical
Specifications (TS) Sections 15.6.2, ``Organization,'' TS 15.6.5.1,
``Manager's Supervisory Staff,'' TS 15.6.6, ``Reportable Event
Action,'' TS 15.6.7, ``Actions To Be Taken If A Safety Limit Is
Exceeded,'' and TS 15.7.8, ``Administrative Controls,'' by changing the
title of the corporate officer responsible for nuclear operations from
the ``Vice President-Nuclear Power,'' to the ``Chief Nuclear Officer.''
Date of issuance: August 25, 1997.
Effective date: August 25, 1997, with full implementation within 45
days.
Amendment Nos.: 177 and 181.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 21, 1997 (62 FR
27802), as corrected May 29, 1997 (62 FR 29163) The August 11, 1997,
submittal provided a corrected TS page. This information was within the
scope of the action noticed and did not change the staff's initial
proposed no significant hazards considerations determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 3rd day of September 1997.
For The Nuclear Regulatory Commission.
Bruce E. Boger,
Director, Division of Reactor Projects--I/II Office of Nuclear Reactor
Regulation.
[FR Doc. 97-23820 Filed 9-9-97; 8:45 am]
BILLING CODE 7590-01-P
