[Federal Register Volume 64, Number 125 (Wednesday, June 30, 1999)]
[Notices]
[Pages 35199-35221]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-16489]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to 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 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 June 5, 1999, through June 18, 1999. The
last
[[Page 35200]]
biweekly notice was published on June 16, 1999 (64 FR 32284).
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 Administration 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 July 30, 1999, 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 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:
[[Page 35201]]
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendments request: May 26, 1999.
Description of amendments request: The proposed amendment would
revise Technical Specification 3.3.1, ``Reactor Protective System (RPS)
Instrumentation--Operating,'' to change the RPS reactor coolant flow
trip setpoints. The change is intended to reduce spurious reactor trip
hazards.
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. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change will change the Reactor Protection
System (RPS) reactor coolant flow trip setpoints. The RPS functions
to mitigate the consequences of an accident. The changes to the low
reactor coolant flow trip setpoints will reduce or eliminate
unnecessary challenges to the RPS. Therefore, the proposed change
will not involve a significant increase in the probability of an
accident previously evaluated.
These changes will result in an increased time delay for the RPS
low reactor coolant flow trip. The reanalysis of the affected UFSAR
[updated final safety analysis report] Chapter 15 events (UFSAR
15.3.4, Reactor Coolant Pump Shaft Break with Loss of Offsite Power
and UFSAR 15.1.5, Steam System Piping Failures Inside and Outside
Containment--Modes 1 and 2 Operations), with the increased time
delay, shows that the dose consequences for these events remain
bounded by the UFSAR analysis. Therefore, this change does not
involve a significant increase in the consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change will change the RPS reactor coolant flow
trip setpoints. The RPS functions to mitigate the consequences of an
accident. The changes to the low reactor coolant flow trip setpoints
will reduce or eliminate unnecessary challenges to the RPS. The
proposed change only changes the mitigating actions of the RPS,
without changing the required function of the RPS. Therefore, the
change to the low reactor coolant flow trip setpoints does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change will change the RPS reactor coolant flow
trip setpoints. The reanalysis of the affected UFSAR Chapter 15
events (UFSAR 15.3.4, Reactor Coolant Pump Shaft Break with Loss of
Offsite Power and UFSAR 15.1.5, Steam System Piping Failures Inside
and Outside Containment--Modes 1 and 2 Operations), with the revised
reactor coolant flow trip setpoints, shows that the minimum DNBR
[departure from nucleate boiling ratio] and SAFDLs [specified
acceptable fuel design limits] for these events remain bounded by
the UFSAR analysis. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Section Chief: Stephen Dembek.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: June 2, 1999.
Description of amendment request: The proposed amendment would
relocate Shearon Harris Nuclear Power Plant (HNP) Technical
Specification (TS) Section 6.5, ``Review and Audit,'' TS 6.8.2, TS
6.8.3, and TS Section 6.10, ``Record Retention,'' intact from the HNP
TS to the Quality Assurance Program Description currently located in
the HNP Final Safety Analysis Report Section 17.3. Future changes to
the associated relocated TS would be processed in accordance with 10
CFR 50.54(a). The proposed change is consistent with NUREG-1431,
Revision 1, ``Standard Technical Specifications, Westinghouse Plants,''
dated April 1995, and with the guidance provided in NRC Administrative
Letter 95-06, ``Relocation of Technical Specification Administrative
Controls related To Quality Assurance,'' dated December 12, 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This TS change relocates administrative requirements from HNP TS
to the Quality Assurance Program Description (QAPD). The proposed
amendment will not introduce any new equipment or require existing
equipment to function different from that previously evaluated in
the Final Safety Analysis Report (FSAR) or TS.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment will not introduce any new equipment or
require existing equipment to function different from that
previously evaluated in the Final Safety Analysis Report (FSAR) or
TS. The changes are consistent with NUREG-1431, Revision 1 and the
Commission's Final Policy Statement on Technical Specification
improvements. The proposed amendment will not create any new
accident scenarios, because the change does not introduce any new
single failures, adverse equipment or material interactions, or
release paths.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
This TS change relocates administrative requirements from HNP TS
to the Quality
[[Page 35202]]
Assurance Program Description (QAPD). The QAPD will be revised to
include the requirements associated with this proposed change. NRC
Administrative Letter 95-06 states that administrative requirements
for review and audit and the independent safety engineering group
may be relocated from TS to the quality assurance program. HNP
proposes relocating the associated requirements from TS to the QAPD
intact. Future changes to these requirements will be processed in
accordance with 10 CFR 50.54(a). This proposed TS change is
administrative in nature and does not alter NRC acceptance limits
with respect to accident mitigation or accident analysis.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Section Chief: Sheri R. Peterson.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 22 and October 22, 1998; May 6,
1999.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) to reflect the licensee's planned use of
fuel supplied by Westinghouse. The staff has published a Notice of
Consideration of Issuance of Amendments and Proposed No Significant
Hazards Consideration Determination on November 3, 1998 (63 FR 69338)
covering the July 22 and October 22, 1998, submittals. In the May 6,
1999, submittal the licensee proposed to expand the original amendment
request, revising Section 5.6.5 of the Technical Specifications.
Section 5.6.5 specifies a list of NRC-approved topical reports that the
licensee is required to use to determine reactor core operating limits.
The licensee proposed to update this list to show the current approval
status of these topical reports.
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 for the proposed changes conveyed by the May 6, 1999,
submittal. The NRC staff has reviewed the licensee's analyses against
the standards of 10 CFR 50.92(c). The NRC staff's analysis is presented
below.
First Standard
No. The proposed changes to Section 5.6.5 will not affect the
safety function, and will not involve any change to the design or
operation of any plant system or component. The topical reports were
previously approved by the NRC staff under separate licensing actions.
The use of methodologies in these approved topical reports will ensure
that previously evaluated accidents remain bounding. Therefore, no
accident probabilities or consequences will be impacted.
Second Standard
No. The proposed changes would not lead to any hardware or
operating procedure change. Hence no new equipment failure modes or
accidents from those previously evaluated will be created.
Third Standard
No. Margin of safety is associated with confidence in the design
and operation of the plant; specifically, the ability of the fission
product barriers to perform their design functions during and following
an accident. The proposed changes to Section 5.6.5 do not involve any
change to plant design, operation, or analysis. Thus the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for the proposed changes to Section 5.6.5.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: April 5, 1999, supplemented May 27,
1999.
Description of amendment request: The proposed amendments would
revise the Improved Technical Specifications (TS), Updated Final Safety
Analysis Report, and Core Operating Limits Report to incorporate
Topical Report (TR) DPC-NE-3005-P, ``Thermal-Hydraulic Transient
Analysis Methodology.'' This analysis has been completed for Unit 2 and
is ongoing for Units 1 and 3. Therefore, the proposed changes that
reflect the TR provisions affect Unit 2 only. Other proposed changes
affect all three units. Specifically, (1) a note to TS Surveillance
Requirement (SR) 3.4.1.2, ``RCS [Reactor Coolant System] Pressure,
Temperature, and Flow DNB [Departure from Nucleate Boiling] Limits,''
would be modified to address application of the delta-Tcold
limits; (2) TS 3.4.10, ``Pressurizer Safety Valves,'' would be modified
to increase the setpoint range of the lift settings for the pressurizer
safety valves for the Oconee unit that has been analyzed in accordance
with the TR and state that the range is not changed for the other
units; (3) a statement to SR 3.4.10.1 would be added that will specify
the pressurizer safety valve lift setpoint in order to clarify the
difference between the operability setpoint range for a test lift and
the range required when the setpoint is reset following the
surveillance test; (4) TS 3.7.4, ``Atmospheric Dump Valve (ADV) Flow
Paths,'' would be added to address the applicability and required
actions related to the ADS valves; (5) TS 3.9.7, ``Unborated Water
Source Isolation Valves,'' would be added to require valves that are
used to isolate unborated water sources to be secured in the closed
position while in Mode 6, incorporate SRs, and provide required actions
if one or more of the valves is not secured in the closed position; (6)
TS 5.6.5b would be changed to update the Core Operating Limits Report
references; and (7) the appropriate Bases would be changed to reflect
the above changes, other changes consistent with the revisions to the
TR analysis, and the Updated Final Safety Analysis Report revisions
that were provided in the submittal.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. The proposed changes to the Technical Specifications, Bases,
Updated Final Safety Analysis Report (UFSAR), and Core Operating
Limits Report (COLR) incorporate the accident analyses established
in Topical
[[Page 35203]]
Report DPC-NE-3005-P, ``UFSAR Chapter 15 Transient Analysis
Methodology.'' On July 30, 1997, Duke submitted Topical Report DPC-
NE-3005-P to the NRC for approval. The NRC found DPC-NE-3005-P
acceptable, with noted exceptions, in a Safety Evaluation issued on
October 1, 1998. To resolve the noted NRC exceptions, Duke submitted
Revision 1 of DPC-NE-3005-P to the NRC for review on February 1,
1999. Additional information regarding Revision 1 of DPC-NE-3005-P
was submitted on April 19 and May 5, 1999. This LAR is dependent
upon the NRC approval of Revision 1 of DPC-NE-3005-P. [This Topical
Report was approved by the NRC on May 25, 1999.]
The analyzed events are initiated by the failure of specific
plant structures, systems or components. These proposed changes do
not impact the condition or performance of those structures, systems
or components.
The revised accident analyses in DPC-NE-3005-P demonstrate that
the applicable acceptance criteria are met. In addition, the
preliminary calculations show that the applicable radiological and
environmental acceptance criteria continue to be met.
Based on the above, the proposed changes do 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 accident previously evaluated?
No. The proposed changes do not involve a physical alteration of
the plant. No new or different equipment is being installed, and no
installed equipment is being operated in a new or different manner.
Where setpoints and operating limits have been revised, the revised
accident analyses demonstrate that the applicable acceptance
criteria are met. As a result, no new failure modes are being
introduced.
Based on the above, the proposed changes do not create the
possibility of any new or different kind of accident from any
accident previously evaluated.
3. Involve a significant reduction in a margin of safety?
No. The margin of safety is established through the design of
the plant structures, systems and components, the parameters within
which the plant is operated, and the establishment of the setpoints
for the actuation of equipment relied upon to respond to a event.
The proposed changes do not involve a physical alteration of the
plant. No new or different equipment is being installed, and no
installed equipment is being operated in a new or different manner.
Where setpoints and operating limits have been revised, the revised
accident analyses in DPC-NE-3005-P demonstrate that the applicable
acceptance criteria are met.
Based on the above, the proposed changes do not involve a
significant reduction in a margin of safety.
Based upon the preceding evaluation, performed pursuant to 10
CFR 50.92, Duke has concluded that the proposed changes to the
Oconee Nuclear Station Technical Specifications, Bases, UFSAR, and
O2C18 COLR will 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: Conee County Library, 501 West
South Broad Street, Walhalla, South Carolina
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: May 24, 1999
Description of amendment request: The proposed amendments would
revise the maximum local fuel pin centerline temperature safety limit
in Technical Specification 2.1.1.1 from the limit determined using the
TACO2 fuel performance computer code to the value determined using a
newer TACO3 computer code.
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 following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92 (c) requirements to demonstrate that all three standards for
no significant hazards consideration are satisfied. A no significant
hazards consideration is indicated 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.
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The use of the revised maximum local fuel pin
centerline temperature limit is appropriate since the new limit uses
a fuel melt temperature which has been conservatively reduced to
account for code uncertainties in calculating fuel centerline
temperature. NRC has previously found the use of the TACO3 code by
DPC [Duke Power Company] in performing reload licensing to be
acceptable. The use of the revised limit for fuel analyzed using an
approved code ensures centerline fuel melting is avoided by ensuring
the maximum fuel temperature is less than the melting temperature of
the fuel. Therefore this change would not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Second Standard
Implementation of this amendment will not create the possibility
of a new or different kind of accident from any previously
evaluated. The use of the revised maximum local fuel pin centerline
temperature limit has no affect on accident precursors.
Implementation of this amendment will not impact any plant systems
that are accident initiators. No other modifications are being
proposed in the plant that would result in the creation of a new
accident mechanism. Also, no changes are being made to the way the
plant is operated; therefore, no new failure mechanisms will be
initiated.
Third Standard
The revised maximum local fuel pin centerline temperature limit
has been appropriately reduced to account for uncertainties in
predicting centerline fuel temperatures. NRC has previously found
the use of the TACO3 code by DPC in performing reload licensing to
be acceptable. Therefore, implementation of this amendment would not
involve a significant reduction in a margin of safety.
Therefore, Duke has concluded that the proposed amendment does
not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of amendment request: May 27, 1999.
Description of amendment request: The proposed changes would
relocate the seismic monitoring instrumentation requirements contained
in Technical Specification (TS) 3/4.3.3.3 to the Licensing Requirements
Manual based on the guidance provided in Generic Letter 95-10,
``Relocation of Selected Technical Specifications Requirements
[[Page 35204]]
Related to Instrumentation.'' The Bases section for Specification 3/
4.3.3.3 will also be relocated to the LRM. The appropriate Index pages,
Table Index page (Unit No. 1 only), TS pages and Bases pages will be
revised to reflect the removal of the seismic monitoring
instrumentation specification from the TSs. An additional specification
page will be added to reflect that Specification Number 3/4.3.3.4 is
not used. This additional page will also denote the number of the
following page. The Bases section will also be modified to denote that
Specification Number 3/4.3.3.4 is not 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:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment would relocate Technical Specification
(TS) 3/4.3.3.3 titled ``Seismic Instrumentation'' and the associated
Bases section to the Licensing Requirements Manual (LRM) (based on
the guidance provided in Generic Letter (GL) 95-10, ``Relocation of
Selected Technical Specification Requirements Related to
Instrumentation''). The proposed amendment would also revise the TS
Index and Beaver Valley Power Station (BVPS) Unit No. 1 List of
Tables to reflect the relocation of this TS and associated Bases.
The relocated Specification will be controlled in accordance with
the requirement of 10 CFR 50.59, ``Controls, Tests, and
Experiments.'' Additional administrative changes are also included
to reflect that Specification Number 3/4.3.3.4 is not used.
The proposed amendment does not involve a significant increase
in the probability of an accident previously evaluated because no
changes are being made to any accident initiator. No analyzed
accident scenario is being changed. The initiating condition and
assumptions remain as previously analyzed. The failure of the
seismic monitoring instrumentation to detect a seismic event is not
an accident initiating event.
The seismic monitoring instrumentation performs no role in
mitigating a seismic event or in achieving a safe shutdown condition
after a seismic event has occurred. Seismic instrumentation is not
assumed to function in the safety analysis. The seismic
instrumentation is not associated with a process variable, design
feature, or operating restriction that is an initial condition of a
Design Basis Accident (DBA) or transient that either assumes the
failure of or presents a challenge to the integrity of a fission
product barrier. Seismic instrumentation does not actuate any
protective equipment or play any direct role in the mitigation of an
accident. The capability of the plant to withstand a seismic event
or other design basis accident is determined by the initial design
and construction of systems, structures, and components. This
instrumentation is used to alert operators to the seismic event and
evaluate the plant response.
The proposed revisions to the Index pages, Table Index page
(BVPS Unit No. 1 only), Specification pages and Bases pages are
administrative in nature and do not affect plant safety.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment does not involve any physical changes to
the plant or the modes of plant operation defined in Appendix A of
the operating license. The proposed amendment does not involve the
addition or modification of plant equipment nor does it alter the
design or operation of plant systems. Seismic instrumentation does
not actuate any protective equipment or play any direct role in the
mitigation of an accident. The capability of the plant to withstand
a seismic event or other design basis accident is determined by the
design and construction of systems, structures, and components. This
instrumentation is used to alert operators to the seismic event and
evaluate the plant response.
Therefore, operation of the facility in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed amendment does not involve revisions to any safety
limits or safety system setting that would adversely impact plant
safety. The proposed amendment does not affect the ability of
systems, structures or components important to ensure the safe
shutdown of the facility, or the mitigation and control of accident
conditions within the facility. In addition, the proposed amendment
does not affect the ability of safety systems to ensure that the
facility can be maintained in a shutdown or refueling condition for
extended periods of time, or the availability of sufficient
instrumentation and control capability for monitoring and
maintaining the unit status.
The proposed revisions to the Index pages, Table Index page
(BVPS Unit No. 1 only), Specification pages and Bases pages are
administrative in nature and do not affect plant safety.
Therefore, the proposed amendment 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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of amendment request: May 27, 1999.
Description of amendment request: The proposed amendments would (1)
revise the frequency for performing the CHANNEL FUNCTIONAL TEST (CFT)
of the manual initiation functional units specified in the Beaver
Valley Power Station, Unit Nos. 1 and 2, Engineered Safety Features
Actuation System (ESFAS) Instrumentation Technical Specifications (TSs)
from monthly, with an accompanying footnote which allows the manual
initiation to be tested on a refueling interval, to each refueling
interval; (2) Revise footnotes associated with TS ESFAS tables; (3)
revise associated TS Bases.
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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change revises the frequency notation specified for
the channel functional test of the manual initiation functions
listed on Table 4.3-2 of TS 3/4.3.2, ``Engineered Safety Feature
Actuation System (ESFAS) Instrumentation.'' The proposed change
revises the current TS requirement for surveillance testing these
functions to clarify that testing be performed on a refueling basis.
The revision to the surveillance frequency specified in Table 4.3-2
does not physically impact the Instrumentation, its setpoints, or
the actual frequency at which the manual initiation functions are
tested. The revision eliminates the potential for confusion
regarding the testing required for the manual initiation function by
deleting Footnote (1) to Table 4.3-2. The proposed change to the
Surveillance Requirements of Table 4.3-2 for the manual initiation
functions eliminates the need for Footnote (1). Footnote (1)
requires testing the manual actuation switches every 18 months and
performing a Channel Functional Test on all other circuitry
associated with manual safeguards actuation every 31 days. As there
is no other circuitry for which a 31 day CFT is applicable, the
proposed change simplifies the TS requirement consistent with the
current Standard TS for Westinghouse plants. Footnote (1) is
consistent with early versions of the Standard Technical
Specifications of
[[Page 35205]]
NUREG-0452; however, later versions of the Standard Technical
Specifications and the Improved Standard Technical Specifications of
NUREG-1431 simply require testing manual initiation functions on a
refueling or 18 month basis. The proposed refueling frequency for
testing this instrumentation recognizes that the manual initiation
functions can not be tested at power since this would introduce the
potential for a significant plant transient.
The deletion of Table 4.3-2 Footnote (1) resulted in renumbering
Footnote (2) to (1). In addition, expired Unit 2 Table 4.3-2
Footnote (3) (only applicable to the first refueling outage) was
also deleted. In addition, changes to the TS bases are made to
further clarify the channel functional test requirements. The
reorganization of the Table 4.3-2 footnotes and bases modifications
are considered to be editorial changes.
The manual initiation instrumentation will continue to be tested
in the same manner as before (every refueling). This test frequency
is consistent with the licensing basis for testing this
instrumentation described in the Updated Final Safety Analysis
Report (UFSAR) and with the testing frequency specified in the
standard Westinghouse Plant TS. Therefore, this test frequency is
considered adequate to verify instrumentation operability. In
addition, failure of a manual initiation function is not an accident
initiator. As such, the ESFAS instrumentation will continue to be
capable of providing the required safety functions described in the
UFSAR. Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequence of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
There are no hardware changes associated with this license
amendment nor are there any changes in the method by which any
safety-related plant system performs its safety function. No new
accident scenarios, transient precursors, failure mechanisms or
limiting single failures are introduced as a result of these
changes. These changes do not introduce any adverse effects or
challenges to any safety-related systems. No change is required to
any system configurations, plant equipment or analyses. Therefore,
these changes will not create the possibility of any new or
different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety depends on the maintenance of specific
operating parameters and systems within design requirements.
Updating the manual initiation function surveillance interval
requirements specified on ESFAS TS Table 4.3-2 and deleting Table
4.3-2 Footnote (1) reflects the standard Westinghouse Plant TS
requirements for this instrumentation and is consistent with the
design and operation of the plant as described in the UFSAR. In
addition, the proposed change does not reduce the current refueling
interval testing performed on this instrumentation. The refueling
test frequency specified for this instrumentation is consistent with
industry standards and considered adequate to ensure the affected
manual initiation functions are maintained operable. The proposed
change will improve the clarity of the TS requirement by eliminating
the potential for confusion as to when the surveillances are
required to be performed. As such, the proposed change continues to
ensure that the operation of the affected instrumentation is
maintained within its design requirements and that it continues to
be capable of providing the required safety functions described in
the UFSAR. Therefore, operation of the facility in accordance with
the proposed amendment will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: June 1, 1999.
Description of amendment request: The proposed amendment would
revise the surveillance requirements and applicable Bases relevant to
inservice inspection requirements for the portions of the once-through
steam generator (OTSG) tubes adjacent to the primary cladding region of
the upper and lower OTSG tubesheets.
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--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The OTSGs are used to remove heat from the reactor coolant
system during normal operation and during accident conditions. The
OTSG tubing forms a substantial portion of the reactor coolant
pressure boundary. An OTSG tube failure is a breach of the reactor
coolant pressure boundary and is a specific accident analyzed in the
Arkansas Nuclear One, Unit 1 (ANO-1), Safety Analysis Report (SAR).
The purpose of the periodic surveillance performed on the OTSGs
in accordance with ANO-1 Technical Specification (TS) 4.18 is to
ensure that the structural integrity of this portion of the reactor
coolant system will be maintained. The TS plugging limit of 40% of
the nominal tube wall thickness requires tubes to be repaired or
removed from service because the tube may become unserviceable prior
to the next inspection. Unserviceable is defined in the TS as the
condition of a tube if it leaks or contains a defect large enough to
affect its structural integrity in the event of an operating basis
earthquake, a loss-of-coolant accident, or a steam line or feedwater
line break. The proposed TS change allows OTSG tubes with axial TEC
[tube end cracking] indications that do not extend from the cladding
region into the carbon steel interface within the tube-to-tubesheet
rolled joint of the tubesheets to remain in service with existing
degradation exceeding the existing 40% through-wall (TW) plugging
limit.
Extensive testing and plant experience has illustrated that TEC
flaws confined to this area within the OTSG will not result in tube
burst or significant tube leakage under MSLB [main steamline break]
conditions. Potential leakage from tubes with TEC will be bounded by
the MSLB evaluation presented in the SAR. Therefore, allowing TEC
flaws in this specific region to remain in service will not alter
the conditions assumed in the current ANO-1 accident analysis for
OTSG tube failures under postulated accident conditions. In
addition, the condition of the OTSG tubes in this region are
monitored during regular inspection intervals to assess for evidence
of growth. Any growth noted will be addressed through the
operational assessment. Therefore, Entergy Operations has determined
that the identification, monitoring, assessment, and corrective
action programs * * * [associated with the proposed changes]
sufficiently support this change request.
Application of the TEC alternate repair criteria will allow
leaving tubes with TEC indications found in the defined area of the
tubesheets in service while ensuring safe operation by monitoring
and assessing the present and future conditions of the tubes.
Through the inspection, monitoring, and assessment programs
previously mentioned, and the on-line leak detection capabilities
available during plant operation, continued safe operation of ANO-1
is reasonably assured.
Therefore, the application of the TEC alternate repair criteria
* * * does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The implementation of the TEC alternate repair criteria will not
result in any failure mode not previously analyzed. The OTSGs are
passive components. The intent of the TS surveillance requirements
are being met by these proposed changes in that adequate structural
integrity will be maintained. Potential leakage under MSLB
conditions will remain bounded by the current SAR analysis.
Additionally, the proposed change does not introduce any new modes
of plant operation.
[[Page 35206]]
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
The application of an alternate repair criteria for TEC provides
adequate assurance with margin that ANO-1 steam generator tubes will
retain their structural integrity under normal and accident
conditions. The structural requirements of TEC affected tubes have
been evaluated satisfactorily and meet or exceed regulatory
requirements. The tubing region where TEC occurs is constrained
within the tubesheet bore; therefore, there is no additional risk
associated with tube rupture. Main steam line break leakage rates
for these tubes are reasonably assured to remain within the
assumptions of the accident analysis by proper application of the
TEC alternate repair criteria program. Because no appreciable impact
is evidenced on the tubes structural integrity or its potential
leakage rate, the margin to safety remains unaltered.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
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: June 1, 1999.
Description of amendment request: The amendments would revise the
St. Lucie, Units 1 and 2, Technical Specifications (TS), Sections
3.5.2, to allow up to 7 days to restore an inoperable Low Pressure
Safety Injection System train to operable status. The amendments would
also revise the associated surveillance requirements and TS Bases
sections to be consistent with the revisions to TS Section 3.5.2. Minor
editorial changes for the specified Recirculation Actuation Signal
(RAS) verification test are also included to ensure the terminology
used in the specification is consistent with plant design.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments for St. Lucie Plant, Units 1 and 2 will
extend the action completion/allowed outage time (AOT) for a single
inoperable Low Pressure Safety Injection (LPSI) train from 72 hours
to 7 days. A LPSI train is designed as a part of each Emergency Core
Cooling System (ECCS) subsystem to supplement Safety Injection Tank
(SIT) inventory during the early stages of mitigating a Design Basis
Accident. As such, components of the LPSI system are not accident
initiators, and an extended AOT to restore operability of an
inoperable LPSI train would not increase the probability of
occurrence of accidents previously analyzed.
The safety analyses for both St. Lucie Units demonstrate that
ECCS performance acceptance criteria are satisfied with only one of
the two redundant ECCS subsystems operating during the postulated
Design Basis Accident. The proposed technical specification
revisions involve the AOT for a single inoperable LPSI train, and do
not change the conditions assumed for the minimum amount of
operating equipment needed for accident mitigation. Therefore, the
consequences of an accident previously evaluated will not be
significantly increased.
In addition to the preceding evaluation, a Probabilistic Safety
Analysis (PSA) was performed to quantitatively assess the risk
impact of the proposed amendments. It was concluded from the results
of that assessment that the risk contribution of the AOT extension
is very small, and that the net impact of the proposed amendment can
be risk beneficial.
The editorial corrections proposed for the specified RAS
verification test do not alter existing test requirements and have
no impact on the accident analyses. Therefore, operation of either
facility in accordance with its proposed amendment 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 amendments will not change the physical plant or
the modes of plant operation defined in either Facility License. The
changes do not involve the addition or modification of equipment nor
do they alter the design of plant systems. Therefore, operation of
either facility in accordance with its 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.
The margin of safety associated with the ECCS system is
established by acceptance criteria for system performance defined in
10 CFR 50.46. The proposed amendments will not change these
acceptance criteria or the operability requirements for equipment
that is used to achieve such performance as demonstrated in the
plant safety analyses. Moreover, an integrated assessment of the
risk impact of extending the AOT for a single inoperable LPSI train
has concluded that the risk contribution is very small, LPSI system
reliability can potentially be improved, and the net impact of the
proposed change can be risk beneficial. The editorial corrections
proposed for the specified RAS verification test do not alter
existing test requirements and have no impact on the accident
analyses. Therefore, operation of either facility in accordance with
its proposed amendment would not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Sheri R. Peterson.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: May 13, 1999.
Description of amendment request: The proposed amendment would make
changes to the TMI-1 Facility Operating License No. DPR-50 Sections
2.a, 2.c.(3), and 2.c.(7) to delete obsolete or outdated portions of
the license conditions, and would change the Bases for Technical
Specification 3.1.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated. Most of the proposed amendment is only administrative; it
adds to the Technical Specifications generic references to various
documents. These changes have no affect upon the plant design or
operation.
[[Page 35207]]
The proposed change to the Technical Specification Bases 3.1.1
is the removal of the specified pressurizer code safety valve flow-
rate for which no basis could be found and the acceptance of a 3%
setpoint drift (as-found) as per the ASME code. The 3% code limit is
in accordance with the plant's Inservice Test Program submittal,
which was evaluated by the NRC staff for the current 10 year
interval and documented under NRC TAC No. M93777. The [c]orrect
pressurizer code safety valve flow is provided in the FSAR Table
4.2-8. The proposed change is supported by a revise[d] Startup
Accident analysis with the revised safety valve flow-rate at the 3%
setpoint drift, which demonstrated that the acceptance criteria for
the event were met with considerable margin. The proposed change
does not affect the Technical Specification 3.1.1.a, pressurizer
code safety valve operable (as-left) requirement of [plus or minus]
1%.
Therefore, operation in accordance with the proposed amendment
does 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, because no
new failure modes are created by the proposed changes. The
administrative changes are cosmetic and have no impact on plant
design or operation.
3. Operation of the facility in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety. The proposed amendment does not change any operating limits
for reactor operation.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: May 26, 1999.
Description of amendment request: The proposed amendment would
approve changes to the TMI-1 Updated Final Safety Analysis Report
(UFSAR) which would allow use of the EPRI (Electric Power Research
Institute) Conservative Deterministic Failure Margin (CDFM) methodology
for seismic analysis of the portions of the auxiliary steam line
located in the Auxiliary, Control and Fuel Handling buildings at TMI-1.
The licensee determined that these changes to the UFSAR required prior
NRC approval in accordance with 10 CFR 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment, use of CDFM methodology for the
analysis of the auxiliary steam system piping, would not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The analysis of the auxiliary steam pipe using the CDFM
methodology demonstrates that the pipe wall will maintain integrity
sufficient to prevent adverse impact on safety related equipment
during a safe shutdown earthquake (SSE). The methodology is based on
actual earthquake experience data and has been shown to be adequate
to demonstrate that piping systems will maintain integrity. The CDFM
methodology was developed by experts in the field of seismic
analysis and is based on actual earthquake experience and the
results of dynamic tests with large seismic accelerations. The
methodology provides a conservative mechanism for analytically
predicting performance during actual earthquakes, and thus its
application would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment, use of CDFM methodology for the
analysis of the auxiliary steam system piping, would not create the
possibility of a new or different kind of accident from any accident
previously evaluateed.
No changes to plant systems, structures or components are
proposed and no changes to methods of operation [of the plant] are
involved.
3. The proposed amendment, use of CDFM methodology for the
analysis of the auxiliary steam system piping, would not involve a
significant reduction in a margin of safety.
No changes are proposed to operating limits or safety system
settings, or to accident analysis acceptance criteria. The CDFM
methodology provides a conservative mechanism for analytically
predicting system performance during actual earthquakes. Its
application to the auxiliary steam system piping would 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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: June 4, 1999.
Description of amendment request: The amendment revises decay heat
removal capability requirements to ensure that at least two active
methods of decay heat removal capability will be available during
shutdown conditions except when the reactor vessel head is removed and
the fuel transfer canal water level is greater than or equal to 23 feet
above the reactor vessel flange.
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:
GPU Nuclear has determined that this Technical Specification
Change Request poses no significant hazards as defined by NRC in 10
CFR 50.92. 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 because the
proposed changes would remove exceptions for decay heat removal
system operability requirements during the time the plant is in a
Refueling Shutdown with the RCS loop not filled. The proposed
changes effectively add requirements to maintain redundancy in decay
heat removal systems.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because the proposed changes
would not introduce any new failure modes or modify existing
systems.
3. Involve a significant reduction in a margin of safety because
the proposed amendment would not involve changes to the safety
limits, limiting safety system settings, or operating limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 35208]]
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone
Nuclear Power Station, Unit No. 1, New London County, Connecticut
Date of amendment request: April 19, 1999.
Description of amendment request: The proposed amendment would
replace the current set of technical specifications for the Millstone
Unit 1 plant with a new set of technical specifications for the
permanently shutdown status of the plant.
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, a summary of which is presented below:
The proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This proposed change is consistent with the STS [standard
technical specifications]. The relocation of requirements from the
MP1 TS [Millstone Unit 1 Technical Specifications] to the licensee
controlled documents is consistent with the criteria set forth in 10
CFR 50.36 for the content of Technical Specifications. The removal
of definitions, generic LCO [limiting condition for operation]
actions and generic surveillance requirements has no impact on
facility SSCs [structure, system, and components] or the methods of
operation of such SSCs. The deletion of design features and safety
limits not applicable to the permanently shutdown and defueled
status of MP1 has no impact on the remaining DBA [design-basis
accident], the fuel handling accidents in the fuel storage pool. The
removal of LCOs and surveillance requirements which are related only
to the operation of the nuclear reactor or only to the prevention,
diagnosis or mitigation of reactor-related transients or accidents
do not affect the applicable DBA previously evaluated. The critical
safety functions involving core reactivity control, reactor heat
removal, reactor coolant system inventory control and containment
integrity are no longer necessary at MP1. The proposed accidents
involving damage to the reactor coolant system, main steam lines,
reactor core, and the subsequent release of radioactive material are
no longer possible at MP1. Fuel pool cooling and makeup related
equipment and support equipment (e.g., electrical power systems) are
not required to be continuously available since recent analysis
demonstrated that there is up to ten days before fuel storage pool
boiling to effect repairs, establish alternate sources of make up
flow, or establish steady state natural air circulation cooling of
the Reactor Building atmosphere and fuel storage pool water in the
event of a loss of cooling and makeup flow to the fuel pool. The
radioactive decay of the irradiated fuel since shutdown of the
reactor in November, 1995 has reduced the consequences of the fuel
handling accident to levels well below those previously analyzed.
The relevant parameter (water level) associated with the fuel pool
provides an initial condition for the fuel handling accident
analyses and is included in the PDTS [Permanently Defueled Technical
Specifications]. The Reactor Building crane LCOs are retained to
preserve the engineered controls which preclude a spent fuel cask
drop from occurring over the fuel storage pool. The deletion and
modification of provisions of the administrative controls do not
directly affect the design of SSCs necessary for safe storage of
irradiated fuel or the methods used for handling and storage of such
fuel in the fuel pool. The relocation of administrative controls
related to quality assurance to the Northeast Utilities Quality
Assurance Program is also consistent with the guidance provided in
NRC Administrative Letter AL 95-06, ``Relocation of Technical
Specification Administrative Controls Related to Quality
Assurance,'' dated December 12, 1995. The changes to the
administrative controls are administrative in nature and do not
affect any accidents applicable to the safe storage of irradiated
fuel or the permanently shutdown and defueled condition of the
reactor. Therefore, the proposed changes to the MP1 TS do not
involve any increase in the probability or consequences of any
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes have no impact on facility SSCs affecting
the safe storage of irradiated fuel or on the methods of operation
of such SSCs, or handling and storage of such fuel. These changes
are consistent with the STS and add to the clarity and ease of use
of the proposed PDTS. The removal of Technical Specifications which
are related only to the operation of the nuclear reactor or only to
the prevention, diagnosis, or mitigation of reactor-related
transients or accidents cannot result in different or more adverse
failure modes or accidents than previously evaluated because the
reactor is permanently shutdown and defueled and MP1 is no longer
authorized to operate the plant. The proposed deletion of provisions
of the MP1 TS do not affect systems credited in the accident
analyses for the fuel handling accident in the fuel storage pool at
MP1. The proposed PDTS continue to require proper control and
monitoring of safety significant parameters and activities. The
proposed restriction on the fuel pool level is fulfilled by normal
operating conditions and preserves initial conditions assumed in the
analyses of the postulated DBA. Reactor Building crane LCOs are
retained from current Technical Specifications to preclude the
possibility of a spent fuel cask drop over the fuel storage pool.
Therefore, the proposed changes to this section of the MP1 TS would
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 deletion of provisions of the MP1 TS, which are not related
to the storage of irradiated fuel or which are inconsistent with the
scope of the STS, will not affect the analyses of the remaining DBA
applicable to MP1. The postulated DBAs involving the reactor are no
longer possible due to the permanently shutdown and defueled
condition of the reactor. The requirements for SSCs which have been
deleted from the MP1 TS are not credited in the existing accident
analyses for the remaining applicable postulated accidents and
therefore, do not contribute to the margin of safety associated with
the accident analysis. Therefore, the proposed changes to this
section of the MP1 TS do not involve any reduction in a margin of
safety.
Conclusion
NNECO has concluded that the proposed change to the MP1 Technical
Specifications does not involve a significant hazards consideration as
defined by 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: Michael T. Masnik.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: January 25, 1996, as supplemented April
26, 1996, September 12, 1996, March 17, 1997, September 9, 1997,
December 30, 1998, and May 19, 1999.
Description of amendment request: The proposed changes extend the
allowed outage time for an emergency diesel generator (EDG) system from
7 to
[[Page 35209]]
14 days. At FitzPatrick, an EDG system consists of 2 EDGs powering one
of two emergency AC power buses. The proposal includes provisions for a
Configuration Risk Management Program (CRMP) consistent with the
guidance of Regulatory Guide (RG) 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications.'' The
NRC staff had previously published a notice on these topics on March
27, 1996 (61 FR 13532). This revised notice on these topics is required
to address revisions made in the licensee's supplemental submittals.
The licensee's January 25, 1996, submittal also proposed two line-
item changes to reduce EDG testing at power and to revise AC power
requirements for cold shutdown and refueling modes. The two line-item
changes have not been affected by the supplemental information provided
by the licensee, so the March 27, 1996, proposed finding of no
significant hazards considerations remains valid for these items.
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:
Operation of the FitzPatrick plant in accordance with the
additional changes to the proposed Amendment discussed above, would not
involve a significant hazards consideration as defined in 10 CFR 50.92,
since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to the Technical Specifications will allow
longer Allowed Out of Service Times to perform necessary repair and
maintenance on Emergency Diesel Generators while at power. This
extended AOT [allowed outage time] will enhance scheduling of
preventive maintenance of individual EDGs without significantly
increasing the probability or consequences of an accident previously
evaluated. The risk evaluations for the EDGs determined that the
probability of an accident by increasing the AOT for an EDG System
from 7 days to 14 days is non-risk-significant.
Increasing the EDG AOT does not involve physical alteration of
any plant equipment and does not affect analysis assumptions
regarding functioning of required equipment designed to mitigate the
consequences of accidents. Further, the severity of postulated
accidents and resulting radiological effluent releases will not be
affected by the increased AOT for an EDG System.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[The CRMP provides administrative controls to ensure equipment
configurations do not result in any significant increase in plant
risk. In RG 1.177, the NRC staff established a standard for the
content of the CRMP. The licensee's proposal is consistent with that
standard, and so 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 accident previously evaluated.
Extending the AOT for an EDG system does not necessitate
physical alteration of the plant or changes in parameters governing
normal plant operation. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated for [the] JAF [FitzPatrick] plant.
[The CRMP provides administrative controls to ensure equipment
configurations do not result in any significant increase in plant
risk. These administrative controls do not create any new equipment
configurations, or provide for operation of equipment in a new or
different manner. Therefore, the CRMP does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.]
3. Involve a significant reduction in the margin of safety.
As discussed above, a Fitzpatrick evaluation determined that the
change in risk associated with extending the AOT for a[n] EDG System
is non-risk-significant. In addition, the design provides adequate
redundancy for safe shut down during the AOT with an EDG System out
of service. This is supported by the LOCA [loss-of-coolant accident]
analyses including analyses for long term suppression pool cooling
and reactor shutdown cooling.
[The CRMP provides administrative controls to ensure equipment
configurations do not result in any significant increase in plant
risk. These administrative controls do not create any new equipment
configurations, or provide for operation of equipment in a new or
different manner. Therefore, the proposed CRMP does not involve a
significant reduction in the margin of safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: May 24, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to correct typographical and
editorial errors, and is considered administrative in nature.
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 editorial and administrative changes involve
typographical errors and/or reflect changes that were previously
reviewed and approved by the NRC. These changes, therefore, do not
modify or add any initiating parameters that would significantly
increase the probability or consequences of any previously analyzed
accident.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
These proposed changes do not involve any potential initiating
events that would create the possibility of a new or different kind
of accident. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
These changes are editorial in nature and/or reflect information
previously reviewed and approved by the NRC. The proposed changes
will make the information in the TS consistent with that already
approved by the NRC. 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, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear
Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
[[Page 35210]]
Sacramento Municipal Utility District (the District), Docket No. 50-
312, Rancho Seco Nuclear Station, Sacramento County, California
Date of amendment request: April 23, 1999.
Description of amendment request: The proposed amendment would
change Permanently Defueled Technical Specification (PDTS) D3/4.1,
``Spent Fuel Pool Level,'' to replace a specific reference to spent
fuel pool (SFP) level alarm switches with a generic reference to SFP
level instrumentation. This would allow the licensee to replace the old
level alarm switches with a new ultrasonic level transmitter.
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:
PA-193 will not create a significant increase in the probability
or consequences of an accident previously evaluated in the SAR
[Safety Analysis Report], because the proposed PDTS change is
editorial in nature and only changes the type of equipment that is
referenced in surveillance specification D4.1.2. The SFP level
instrument reference in D4.1.2 is changed from a specific reference
(i.e., SFP level alarm switches) to a more generic reference (i.e.,
SFP level instrumentation). In addition:
1. SFP level monitoring instrumentation is not relied on to
mitigate the consequences of the accidents analyzed in the SAR
(i.e., Fuel Handling Accident, Loss-Of-Offsite-Power event, Liquid
Tank Ruptures, and Decommissioning Accidents),
2. PA-193 does not alter the SFP level monitoring, SFP cooling,
or fuel handling functions during the PDM [Permanently Defueled
Mode],
3. PA-193 continues to require an 18-month calibration of SFP
level instrumentation, and
4. SFP level and alarm indication in the Control Room is
maintained with the new SFP level instrumentation. Also, the SFP
level alarm setpoints remain unchanged with the new SFP level
detection system.
PA-193 will not create the possibility of a new or different
type of accident than previously evaluated in the SAR, because SFP
level instrumentation does not provide any control function and does
not affect any equipment associated with SFP cooling, fuel handling,
or inventory control. The proposed wording change to PDTS D4.1.2
accommodates upgrading the SFP level instrumentation without
changing the intent of surveillance specification D4.1.2. Also, the
new SFP level detection system will (1) maintain the existing SFP
level alarm setpoints and Control Room indication features and (2)
have no adverse impact on the SFP level monitoring function.
PA-193 will not involve a significant reduction in the margin of
safety, because the proposed PDTS change is editorial in nature and
necessary and only accommodates replacing an unreliable, antiquated
SFP level monitoring system with a new, state-of-the-art, ultrasonic
level detection system. The new SFP level detection system will
improve the accuracy, reliability, and serviceability of the SFP
level monitoring function. The District is maintaining the
requirement to perform a[n] SFP level calibration and is only
changing the type of equipment that is referenced in D4.1.2 from a
specific reference (i.e., SFP level alarm switches) to a more
generic reference (i.e., SFP level instrumentation).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Local Public Document Room location: Central Library, Government
Documents, 828 I Street, Sacramento, California 95814.
Attorney for licensee: Dana Appling, Esq., Sacramento Municipal
Utility District, P.O. Box 15830, Sacramento, California 95852-1830.
NRC Section Chief: Michael T. Masnik.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: June 8, 1999 (PCN-495).
Description of amendment requests: The licensee has re-evaluated
its small break loss-of-coolant accident (SBLOCA) using ABB Combustion
Engineering (ABB-CE) S2M evaluation model. Based on this re-evaluation,
the licensee proposes to revise the Technical Specifications (TSs) for
the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3 to
reflect that charging flow is not required to mitigate the effects of
the SBLOCA, add a surveillance requirement to verify that each charging
pump is operable for boration based on the Inservice Testing Program,
increase the maximum as-found lift pressure positive tolerance of main
steam safety valves (MSSVs) from +1% to +2% of the lift setting, and
list the ABB-CE S2M model as an acceptable method for determining
linear heat rate. The licensee will also revise the TS Bases and the
Updated Final Safety Analysis Report (UFSAR) to reflect the proposed
changes.
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) Involve a significant increase in the probability or
consequences of any accident previously evaluated?
Response: No.
The new Small Break Loss Of Coolant Accident (SBLOCA) evaluation
model (ABB Combustion Engineering (ABB-CE) S2M SBLOCA evaluation
model, CENPD 137 Supplement 2-P-A, ``Calculative Methods of the ABB-
CE Small Break LOCA Evaluation Model,'' dated April 1998) more
accurately models the heat transfer mechanisms that occur during a
SBLOCA. As a result of this modeling improvement, there is no longer
a need to credit charging flow during a SBLOCA. The reanalysis, with
an as-found tolerance of +2%/-3% of the lift setting on Main Steam
Safety Valves (MSSVs) 2(3)-PSV-8401 and 2(3)-PSV-8410 in Table
3.7.1-2, determined that the peak cladding temperature (PCT) that
occurs in a SBLOCA is within the acceptance criteria limit of 2200
[degrees] F specified in 10CFR50.46.
This proposed change removes the charging pump Emergency Core
Cooling System (ECCS) surveillance requirement from the Technical
Specifications (TS) which effectively removes the charging system
from the ECCS. This is based on the SBLOCA reanalysis using the new
ABB-CE S2M SBLOCA evaluation model. The reanalysis using the new
model did not credit charging system flow to the reactor coolant
system.
Because this proposed change to remove the charging pump ECCS
flow surveillance requirement is based on a reanalysis of the SBLOCA
rather than physical changes to the plant or the way it is operated,
the probability of the SBLOCA is not affected. The results of the
reanalysis demonstrate the consequences of the SBLOCA without
charging flow do not exceed the consequences of the limiting LOCA.
This is based on the fact that the SBLOCA PCT [peak clad
temperature] does not exceed the limiting large break LOCA PCT.
The addition of Surveillance Requirement (SR) 3.1.9.5 to require
the charging pump to be tested in accordance with the Inservice
Testing (IST) program will ensure that the charging pumps remain
capable of performing their emergency boration requirements.
Use of the NRC approved ABB-CE S2M SBLOCA analysis methodology
identified in TS 5.7.1.5 for calculating the core operating limits
further assures that there is no significant increase in the
probability or consequences of any accident.
Therefore, the probability or consequences of any accident
previously evaluated are not increased.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated?
Response: No.
This change does not involve a physical change to the plant, or
a change to the way the plant is operated. The as-left tolerance of
[plus or minus] 1% on MSSVs 2(3)-PSV-8401 and 2(3)-PSV-8410 in Table
3.7.1-2 is not being changed. The charging system will still be
verified capable of meeting its emergency boration requirements.
[[Page 35211]]
Use of the NRC approved ABB-CE S2M SBLOCA analysis methodology
identified in TS 5.7.1.5 for calculating the core operating limits
further assures that there is no increase in the possibility of a
new or different kind of accident from any previously evaluated.
Therefore, the possibility of a new or different kind of accident
from any previously evaluated is not created.
(3) Involve a significant reduction in a margin of safety?
Response: No.
This proposed change to remove the ECCS surveillance requirement
for the charging pumps, and increase the as-found tolerance on MSSVs
2(3)-PSV-8401 and 2(3)-PSV-8410, is based on a SBLOCA reanalysis
using the new ABB-CE S2M SBLOCA evaluation model. The NRC Safety
Evaluation for the ABB-CE S2M evaluation model determined that the
new evaluation model contains sufficient conservatism such that an
adequate margin of safety exists when the S21VI evaluation model is
used. The results of the SBLOCA reanalysis are within the acceptance
criteria specified in 10 CFR 50.46.
Testing of the charging pumps per the Inservice Testing Program,
combined with the existing Technical Specification 3.1.9--``Boration
System--Operating'' surveillance requirements ensure that the
emergency boration requirements remain met without any reduction in
a margin of safety.
Use of the NRC approved S2M ABB-CE SBLOCA analysis methodology
identified in TS 5.7.1.5 for calculating the core operating limits
further assures that there is no significant reduction in any margin
of safety.
Therefore, a significant reduction in margin of safety is not
involved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 7, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 2.2.1, Reactor Trip System (RTS)
Instrumentation Setpoints, and TS 3.3.2, Engineered Safety Features
Actuation System (ESFAS) Instrumentation, and the associated Bases, by
removing the Total Allowance (TA), Sensor Error (S), and Z terms from
the RTS and ESFAS Instrumentation Trip Setpoints Tables. This would
replace the five-column methodology with a two-column methodology that
consists of the trip setpoint and allowable value columns.
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 involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change eliminates the option to evaluate the
equation (Z+R+S [is less than or equal to] TA), within 12 hours,
from Technical Specification 2.2.1, when the trip setpoint is
outside the allowable value limit. The equation established a
threshold for submitting a Licensee Event Report. The change does
not affect the probability of an accident. The evaluation of the
equation is an administrative provision and has no relevance to the
initiation of any analyzed event. The consequences of an accident
are not affected. The change will not alter assumptions relative to
the mitigation of an accident or transient event.
The proposed amendment is a programmatic and administrative
change that does not physically alter safety-related systems, nor
does it affect the way in which safety-related systems perform their
functions. Because the design of the facility and system operating
parameters are not being changed, the proposed amendment does not
involve an increase in the probability or consequences of any
accident previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendment is a programmatic and administrative
change that does not physically alter safety-related systems, nor
does it affect the way in which safety-related systems perform their
functions. The changes in methods governing normal plant operation
are consistent with current safety analysis assumptions. The
proposed change eliminates the option to evaluate the equation
(described above) within 12 hours, when the trip setpoint is outside
the allowable limit. Because the design of the facility and system
operating parameters are not being changed, the proposed amendment
does not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change does not involve a significant reduction in
a margin of safety.
The proposed amendment is a programmatic and administrative
change that provides assurance that plant operations continue to be
conducted in a safe manner. As stated above, the proposed amendment
does not physically alter safety-related systems, nor does it affect
the way in which safety-related systems perform their functions. The
proposed change eliminates the option to evaluate the equation
(described above) within 12 hours, when the trip setpoint is outside
the allowable limit.
The margin of safety is not affected by eliminating an
administrative provision in Technical Specifications. The
determination for submitting a Licensee Event Report when a trip
setpoint is outside the allowable value will be performed with the
guidelines of 10CFR50.73. The safety analysis assumptions will still
be maintained, thus, no question of safety exists. Because the
design of the facility and system operating parameters are not being
changed, the proposed amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas
77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority, Docket Nos. 50-260, 50-296, Browns Ferry
Nuclear Power Plant, Units 2 and 3. Limestone County, Alabama
Date of amendment request: March 12, 1997 as supplemented by
letters dated March 30, 1999, April 23, 1999 and June 18, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to extend, from 7 days to 14 days,
the Allowable Outage Time (AOT) applicable to an inoperable emergency
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:
No Significant Hazards Consideration Determination
TVA has concluded that operation of BFN in accordance with the
proposed change to the TS does not involve a significant hazards
consideration. TVA's conclusion is based on it's evaluation, in
accordance with 10 CFR 50.91(a)(1), of the three standards set forth
in 10 CFR 50.92(c).
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The EDGs are designed as backup AC power sources in the event of
loss of off-site
[[Page 35212]]
power. The proposed AOT does not change the conditions, operating
configurations, or minimum amount of operating equipment assumed in
the safety analysis for accident mitigation. No changes are proposed
in the manner in which the EDGs provide plant protection or which
create new modes of plant operation. In addition, a PSA evaluation
concluded that the risk contribution of the AOT extension is non-
risk significant. Therefore, the proposed amendment does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not introduce any new modes of plant
operation or make physical changes to plant systems. Therefore,
extension of the allowable AOT for EDGs does not create the
possibility of a new or different accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
BFN's emergency AC system is designed with sufficient redundancy
such that an EDG may be removed from service for maintenance or
testing. The remaining EDGs are capable of carrying sufficient
electrical loads to satisfy the UFSAR requirements for accident
mitigation or unit safe shutdown.
Increasing the allowable EDG AOT will likely increase EDG
unavailability on the average since it expected that the provision
would occasionally be used to accommodate unplanned major EDG
maintenance. However, a conservative PSA evaluation concluded that
the risk contribution of the AOT extension is non-risk significant.
For the 12-year EDG PM work activity, it is expected that the
proposed TS would actually reduce unavailability since multiple
outages would not be necessary to accomplish the maintenance
activity.
The proposed change does not impact the redundancy or
availability requirements of off-site power supplies or change the
ability of the plant to cope with station blackout events. For these
reasons, the proposed amendment 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: Athens Public Library, 405 E.
South Street, Athens, Alabama.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Drive, ET 10H, Knoxville, Tennessee 37902,
NRC Section Chief: Sheri R. Peterson.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: May 4, 1999, as supplemented by letter
dated June 4, 1999.
Brief description of amendments: The proposed license amendments
would revise the Technical Specifications for CPSES, Units 1 and 2.
Specifically, the changes would revise the surveillance requirements
associated with the plant battery and emergency diesel generators, and
correct miscellaneous editorial errors that resulted from the issuance
of Amendment No. 64. The original application was noticed and published
in the Federal Register on June 2, 1999 (64 FR 29715). The June 4,
1999, supplement provided proposed additional editorial corrections.
The supplemental information is being noticed herein to address the
issue of no significant hazards consideration.
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 consequence of an accident previously evaluated?
(1) Batteries are used to support mitigation of the consequences
of an accident, and are not considered to be an initiator of any
previously analyzed accident. The proposed change would not effect
the design or performance of the batteries. The allowance to perform
the modified performance discharge test in lieu of the service test
at any time is permissible since the test's discharge rate envelopes
the duty cycle of the service test. Therefore, the allowance for
unrestricted substitution of the modified performance discharge test
in lieu of the service discharge test does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) The diesel generators are used to support mitigation of the
consequences of an accident, and are not considered to be an
initiator of any previously analyzed accident. The proposed change
does not affect the accident analysis assumption that the DG reaches
minimum conditions to accept load within 10 seconds. The ability of
the DG to maintain steady state operation within 10 seconds is not
an accident analysis assumption and is primarily used to identify
degradation of governor and voltage regulator performance.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(3) The editorial changes are non-technical and therefore do not
involve a significant increase in the probability or 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?
(1) The allowance for unrestricted substitution of the modified
performance discharge test in lieu of the service discharge test
does not involve any physical alteration to the plant. No new
failure mechanisms will be introduced and the change does not affect
the ability of the batteries to fulfill their safety-related
function. Therefore, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(2) The separation of the DG start surveillance criteria into
those criteria required to be met within 10 seconds, and those
criteria required to be met following achievement of steady state
conditions, does not involve any physical alteration to the plant.
No new failure mechanisms will be introduced and the change does not
affect the ability of the DGs to fulfill their safety-related
function. Therefore, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(3) The editorial changes are non-technical and therefore do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
(1) The allowance for unrestricted substitution of the modified
performance discharge test in lieu of the service discharge test
will not alter any accident analysis assumptions, initial
conditions, or results. Consequently, it does not have any effect on
the margin of safety. Therefore, this change does not involve a
significant reduction in a margin of safety.
(2) The proposed change to delete the requirement to demonstrate
that the DG can achieve and maintain steady state operation within
10 seconds is not an accident analysis assumption. The accident
analysis assumption that the DG reaches minimum conditions to accept
load within 10 seconds is preserved. Consequently, it does not have
any effect on the margin of safety. Therefore, this change does not
involve a significant reduction in a margin of safety.
(3) The editorial changes are non-technical and therefore do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
[[Page 35213]]
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: May 14, 1999.
Brief description of amendments: The proposed license amendments
would change the name of the CPSES licensee from ``Texas Utilities
Electric Company'' to ``TXU Electric Company'' in the Facility
Operating Licenses of CPSES, Units 1 and 2.
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. Does the proposed change involve a significant increase in
the probability or consequence of an accident previously evaluated?
No. This request involves an administrative change only. The
Operating Licenses (OLs) are being changed to reference the new
corporate name of the licensee. No actual plant equipment or
accident analyses will be affected by the proposed change.
Therefore, TU [Texas Utilities] Electric concludes that this request
will have no impact on the possibility of any type of accident,
whether new, different or previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. This request involves an administrative change only. The OLs
are being changed to reference the new corporate name of the
licensee. No actual plant equipment or accident analyses will be
affected by the proposed change and no failure modes not bounded by
previously evaluated accidents will be created. Therefore, TU
Electric concludes that this request will have no impact on the
possibility of any type of accident, whether new, different or
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. Margin of safety is associated with confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. This
request involves an administrative change only. The OLs are being
changed to reference the new corporate name of the licensee. No
actual plant equipment or accident analyses will be affected by the
proposed change. Additionally, the proposed change will not relax
any criteria used to establish safety limits, will not relax any
safety systems settings, or will not relax the bases for any
limiting conditions of operation. Therefore, this request will not
impact 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, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: May 24, 1999.
Brief description of amendments: The proposed license amendments
would remove several cycle-specific parameter limits from the Technical
Specifications (TSs) and add parameter limits to the Core Operating
Limits Report. In addition, the core safety limit curves would be
replaced with safety limits more directly applicable to the fuel and
fuel cladding fission product barriers. The affected TSs are: (1) TS
2.0, ``Safety Limits (SLs)''; (2) TS 3.3.1, ``Reactor Trip System
Instrumentation Setpoints''; (3) TS 3.4.1, ``RCS pressure temperature
and flow from Nucleate Boiling (DNB) Limits''; and (4) TS 5.6.5, ``Core
Operating Limits Report.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes remove cycle-specific parameter limits from
the Technical Specifications, add them to the list of limits
contained in the Core Operating Limits Report (COLR), and revise the
Administrative Controls section of the Technical Specifications. The
proposed changes also insert the original minimum RCS [reactor
coolant system] flow limits into the Technical Specifications. The
changes do not, by themselves, alter any of the parameter limits.
The changes are administrative in nature and have no adverse effect
on the probability of an accident or on the consequences of an
accident previously evaluated. The removal of parameter limits from
the Technical Specifications does not eliminate the requirement to
comply with the parameter limits.
The parameter limits in the COLR may be revised without prior
NRC approval. However, [Technical] Specification 5.6.5c continues to
ensure that the parameter limits are developed using NRC-approved
methodologies and that applicable limits of the safety analyses are
met. While future changes to the COLR parameter limits could result
in event consequences which are either slightly less or slightly
more severe than the consequences for the same event using the
present parameter limits, the differences would not be significant
and would be bounded by the requirement of specification 5.6.5c to
meet the applicable limits of the safety analysis.
Based on the above, addition of the minimum RCS flow limit into
the Technical Specifications, removal of the parameter limits from
the Technical Specifications and the addition of the described
limits in the COLR, thus allowing revision of the parameter limits
without prior NRC approval, has no significant effect on the
probability or 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?
The proposed changes add the minimum RCS flow limit into the
Technical Specifications, remove certain parameter limits from the
Technical Specifications and add these limits to the list of limits
in the COLR, thus removing the requirement for prior NRC approval of
revisions to those parameters. The changes do not add new hardware
or change plant operations and therefore cannot initiate an event
nor cause an analyzed event to progress differently. Thus, the
possibility of a new or different kind of accident is not created.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The margin of safety is the difference between the acceptance
criteria and the associated failure values. The proposed changes do
not affect the failure values for any parameter. Through the
accident analyses, all applicable limits (i.e., relevant event
acceptance criteria as described in the NRC-approved analysis
methodologies) are shown to be satisfied; therefore, there is no
impact on event acceptance criteria. Because neither the failure
values nor the acceptance criteria are affected, the proposed change
has no effect on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
[[Page 35214]]
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: May 5, 1999.
Description of amendment request: The proposed change modifies the
Technical Specifications (TS) to enhance limiting conditions for
operation and surveillance requirements relating to the Standby Liquid
Control (SLC) system and incorporates certain provisions of NRC's rule
on anticipated transients without scram (ATWS) (10CFR50.62). The change
involves the use of enriched boron in the SLC system and improves upon
other aspects of the TS for this 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:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change deletes the requirement for standby liquid
control (SLC) system operability during refueling and modifies the
conditions for allowing the system to be inoperable when shutdown.
This change also permits changing the reactor mode switch to the
``Run'' or ``Startup/Hot Standby'' position to test mode switch
interlock functions while the SLC system is inoperable. To allow
testing of instrumentation associated with the reactor mode switch
interlock functions, compensatory measures are provided for assuring
that no core alterations are in progress and that all control rods
remain fully inserted in core cells containing one or more fuel
assemblies. These compensatory measures ensure that no credible
mechanisms for an inadvertent criticality are introduced by
administratively controlling the required functions of the reactor
mode switch interlocks. Control rods are not required to be inserted
in empty core cells (i.e., those containing no fuel) because, with
one or more cells in this configuration, the overall shutdown margin
is actually greater than when all control rods and all fuel
assemblies are inserted.
The SLC system is not assumed in the initiation of any
previously evaluated events and therefore the proposed change will
not significantly increase the probability or consequences of a
previously analyzed accident. The SLC system is not assumed to
operate in the mitigation of any previously analyzed accidents which
are assumed to occur during shutdown or refueling conditions. This
change will not result in operation that will significantly increase
the probability of initiating an analyzed event. This change will
not alter assumptions relative to mitigation of an accident or alter
the operation of process variables, structures, systems, or
components as described in the final safety analysis report.
VY has determined that the proposed change to increase the
standby liquid control system reactivity control capacity using a
borated water solution enriched in the boron-10 isotope effectively
increases the rate of injection of neutron absorber and does not
alter the function of the system, method of operation or dual train
configuration. The system response time to an anticipated transient
without scram (ATWS) event has been reduced as the increased boron-
10 enrichment of the solution provides faster negative reactivity
insertion, thus reducing the consequences of the ATWS event. The SLC
system is not credited in any of the design basis accident analyses
and, as such, is considered to provide only an additional mitigative
feature in the event of an accident. The SLC system sodium
pentaborate solution concentration and flow rate required by the
ATWS rule (10CFR50.62) for reactivity control independent of the
control rods are not reduced from the values previously evaluated
and presented in the Vermont Yankee Technical Specifications. The
addition of enriched boron provides a shutdown margin greater than
the previously calculated shutdown reactivity control capacity, and
the change does not affect the probability of an ATWS event.
Therefore, this change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment, will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change modifies the modes of applicability for the
SLC system. Included in this change is allowance to permit changing
the reactor mode switch to the ``Run'' or ``Startup/Hot Standby''
position to test mode switch interlock functions while the SLC
system is inoperable. Precautions are taken when manipulating the
mode switch to one of these positions to maintain all control rods
fully inserted in core cells containing at least one fuel assembly
and to not allow any core alterations. These two provisions
eliminate the possibility of introducing any credible mechanisms for
inadvertent criticality. The proposed change will not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or changes in methods governing normal
plant operation. The proposed change will not eliminate any valid
requirements necessary for safe operation.
VY has determined that the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated because the proposed change involves a system
whose function is to provide an additional (backup) mitigative
shutdown capability and no system modifications are made.
The addition of enriched boron does not affect any system or
component that could initiate an accident. Thus, no new or different
type of accident is created.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
VY has determined that the proposed change does not involve a
significant reduction in a margin of safety. The proposed change
would remove the backup to the available reactivity control systems
when the reactor is in a shutdown or refueling condition. However,
this backup is not considered in the margin of safety when
determining the required reactivity for shutdown and refueling
events. This change will have no impact on any safety analysis
assumptions.
Included in this change is allowance to permit changing the
reactor mode switch to the ``Run'' or ``Startup/Hot Standby''
position to test mode switch interlock functions while the SLC
system is inoperable. The margin of safety will not be reduced
during such testing of interlock functions with the SLC system
inoperable because compensatory measures have been added to ensure
that no credible mechanisms for inadvertent criticality exist with
the reactor mode switch in other than the ``Shutdown'' or ``Refuel''
positions.
The use of enriched boron in the SLC system sodium pentaborate
solution actually increases the capability of the SLC system to
achieve cold shutdown; thus, no margin of safety is reduced.
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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: June 3, 1999.
Description of amendment request: The request is to amend the
operating license such that the name of the licensee is changed from
Washington Public Power Supply System to Energy Northwest. The name of
the facility will be changed from WPPS Nuclear Project No. 2 to WNP-2.
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:
[[Page 35215]]
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This request involves an administrative change only. The
Operating License (OL) is being changed to reference the new name of
the licensee. No actual plant equipment or accident analyses will be
affected by the proposed change. Therefore, this request will have
no impact on the probability or consequence of any type of accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This request involves an administrative change only. The OL is
being changed to reference the new name of the licensee. No actual
plant equipment or accident analyses will be affected by the
proposed change and no failure modes not bounded by previously
evaluated accidents will be created. Therefore, this request will
have no impact on the possibility of any new type of accident: new,
different, or previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Margin of safety is associated with the confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. This
request involves an administrative change only. The OL is being
changed to reference the new name of the licensee.
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed change will not
relax any criteria used to establish safety limits, will not relax
any safety system settings, or will not relax the bases for any
limiting conditions of operation. Therefore, this request will not
impact the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 10, 1999.
Description of amendment request: The amendment would revise
Technical Specification Table 3.3-4, Functional Unit 7.b., Automatic
Switchover to Containment Sump (Refueling Water Storage Tank Level--
Low-Low) to reflect the results of calculations that were performed for
the associated instrumentation setpoints to consider the density
variations due to temperature and boric acid concentration.
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 protection system performance will remain within the bounds
of the previously performed accident analysis. The protection
systems will continue to function in a manner consistent with the
plant design basis. The proposed changes will not affect any of the
analysis assumptions for any of the accidents previously evaluated,
since the changes are consistent with the setpoint methodology and
ensure adequate margin to the Safety Analysis Limit. The proposed
changes will not affect any event initiators nor will the proposed
changes affect the ability of any safety related equipment to
perform its intended function. There will be no degradation in the
performance of nor an increase in the number of challenges imposed
on safety related equipment assumed to function during an accident
situation. There will be no change to normal plant operating
parameters or accident mitigation capabilities.
Therefore these changes 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.
There are no changes in the method by which any safety related
plant system performs its safety function. The normal manner of
plant operation remains unchanged, and no new equipment is being
introduced. The increase in the RWST [refueling water storage tank]
Level Low-Low Allowable Value still provides acceptable margin
between the nominal Trip Setpoint and Allowable Value while taking
into account a temperature and boric acid density correction. The
change in Allowable Value does not impact the systems capability to
perform an ECCS [emergency core cooling system] switchover from
injection to cold leg recirculation since the nominal Trip Setpoint
remains the same. The change in Allowable Value also will not affect
injection or recirculation of the Containment Spray System.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed changes. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change in any Safety Analysis
Limit. There will be no effect on the manner in which safety limits
or Engineered Safety Features Actuation System settings are
determined nor will there be any affect on those plant systems
necessary to assure the accomplishment of protection functions.
Therefore, there will be no impact on any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 11, 1999.
Description of amendment request: The amendment would revise
Technical Specification 3.7.1.6, ``Steam Generator Atmospheric Relief
Valves,'' and its associated Bases to (1) require four atmospheric
relief valves (ARVs) to be operable; (2) eliminate the use of
``required'' in the action statements; (3) provide action statements to
address inoperability of two ARVs and three or more ARVs due to causes
other than excessive leakage; and (4) limit the Limiting Condition for
Operation (LCO) 3.0.4 exception to one inoperable ARV.
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.
Revising the LCO to require four ARVs to be OPERABLE rather than
three; eliminating
[[Page 35216]]
``required'' from the Actions; adding a new ACTION for three or more
ARVs inoperable; and limiting the LCO 3.0.4 exception to one ARV
inoperable constitute more restrictive changes from the current
Technical Specifications. The proposed changes do not affect
initiating mechanisms or mitigation capabilities associated with
SGTR [steam generator tube rupture] events analyzed in Chapter 15 of
the Updated Safety Analysis Report. The proposed changes impose more
stringent requirements to ensure that ARV OPERABILITY is maintained
consistent with the safety analysis and licensing basis, and also to
address all potential single failure scenarios. Therefore these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
With two ARVs inoperable, the allowed outage time for
restoration of all but one ARV to OPERABLE status is changed from 24
hours to 72 hours. The existing specification allows one valve to be
inoperable indefinitely and with one required ARV inoperable, the
allowed outage time for restoration is seven days. By modifying the
LCO to require four ARVs to be OPERABLE, an allowed outage time of
72 hours is more restrictive than the existing specification.
Therefore, revising the allowed outage time from 24 hours to 72
hours is acceptable based on a more restrictive allowed outage time
from the existing specification and the low probability of an event
requiring decay heat removal occurring during the restoration period
that would require the ARVs. With respect to Reactor Coolant System
cooldown for SGTR accident mitigation, the increase in time is
acceptable based on the low probability of a SGTR event occurring
during the restoration period and the low probability of a SGTR
event in conjunction with the failure of the turbine bypass system
(i.e., loss of offsite power). Therefore, this change in allowed
outage time does not result in a significant increase in the
probability or consequences of previously analyzed accidents.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. Revising the LCO to require four ARVs to be OPERABLE
rather than three; eliminating ``required'' from the Actions; adding
a new ACTION for three or more ARVs inoperable; and limiting the LCO
3.0.4 exception to one ARV inoperable will not impact the normal
method of plant operation. The proposed changes ensure operation of
the plant remains consistent with analysis assumptions. No new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. Based on the above discussion, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any affect on those plant systems necessary to assure
the accomplishment of protection functions. The proposed changes
ensure operation of the plant consistent with the analysis
assumptions. Therefore, there will be no impact on any margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notice of Consideration of Issuance of
Amendment to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: May 24, 1999.
Description of amendment request: Clarify nonconservative wording
of Technical Specification (TS) 3/4,5,1, ``Safety Injection Tanks,''
and revise TS 3/4.5.2, ``ECCS Subsystems--Tavg Greater Than or Equal to
325 degrees F,'' to align their associated surveillance requirements
with the intent and design bases requirements intended to be verified.
Date of publication of individual notice in the Federal Register:
June 10, 1999 (64 FR 31322).
Expiration date of individual notice: June 25, 1999.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
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.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: April 12, 1999.
Brief description of amendment: The amendment is a temporary
amendment change effective until September 30, 1999, which revises
Technical Specification 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit
an 8-hour delay in the UHS temperature restoration period prior to
entering the plant shutdown required actions.
Date of issuance: June 4, 1999.
Effective date: June 4, 1999.
Amendment No.: 183.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24193).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 4, 1999.
[[Page 35217]]
No significant hazards consideration comments received: No.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear Power
Station, Unit 3, Grundy County, Illinois
Date of application for amendment: May 5, 1999.
Brief description of amendment: The amendment removes the safety
valve function of the Target Rock safety/relief valve from Technical
Specifications (TS) Section 3.6.E and moves the reactor coolant system
safety valve lift pressure setpoints from TS Section 3.6.E to TS
Section 4.6.E.
Date of issuance: June 4, 1999.
Effective date: As of the date of issuance and shall be effective
within 30 days from the date of issuance.
Amendment No.: 168.
Facility Operating License No. DPR-25: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 21, 1999 (64 FR
27824).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: December 7, 1998, as
supplemented May 12, 1999.
Brief description of amendment: The amendment revises Technical
Specification 4.13A.2.a. to allow a one-time extension of the steam
generator (SG) inspection interval. In addition, the amendment would
remove the requirement of receiving NRC concurrence on the proposed SG
examination program in TS 4.13C.1.
Date of issuance: June 9, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 201.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6694).
The May 12, 1999, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 29, 1996, as supplemented January
8, 1998.
Brief description of amendment: The proposed changes revise
requirements prescribed in Technical Specification Surveillance
Requirement 3.3.1.1.8 and allow River Bend to increase the interval
between whole core traversing in-core probe to local power range
monitor calibrations from 1,000 megawatt days per ton (MWD/T) to 2,000
MWD/T.
Date of issuance: June 11, 1999.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 107.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
55032).
The January 8, 1998, letter provided additional information that
did not change the scope of the original application and the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 11, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: April 30, 1998.
Brief description of amendment: The amendment revises the
definition of quadrant power tilt to clearly allow the use of either
the incore detectors or the excore detectors for determining quadrant
power tilt.
Date of issuance: June 10, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 197.
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6694).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: April 9, 1999.
Brief description of amendment: The proposed amendment modifies the
Technical Specifications (TSs) to add Limiting Condition for Operation
3.0.6 and its associated Bases. This change allows equipment that has
been removed from service or declared inoperable in compliance with the
TS Action statement to be returned to service under administrative
controls solely to perform testing required to demonstrate its
operability or the operability of other equipment. The proposed change
is consistent with TS 3.0.5 as discussed in NUREG-1432, Revision 1,
``Standard Technical Specifications for Combustion Engineering
Plants.'' TS 3.0.2 is also modified to reflect that TS 3.0.6 is an
exception to TS 3.0.2.
Date of issuance: June 7, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance: June 7, 1999.
Amendment No.: 207.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24196).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 7, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
[[Page 35218]]
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: October 1, 1998, as supplemented by
letters dated March 25 and May 6, 1999.
Brief description of amendment: The amendment modifies Technical
Specification (TS) 3.3.3.7.3 and Surveillance Requirement 4.3.3.7.3 for
the broad range gas detection system at Waterford 3. In addition, TS
Bases 3/4.3.3.7 has been changed to reflect the new system.
Date of issuance: June 3, 1999.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 151.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64114).
The March 25 and May 6, 1999, letters provided clarifying
information that did not change the scope of the original application
and the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 25, 1999, as supplemented by
letter dated April 16, 1999.
Brief description of amendment: The amendment removes certain
administrative controls from the Waterford 3 Technical Specifications
and instead relies on the requirements of the new Entergy common
Quality Assurance Program Manual and the change controls of Title 10 of
the Code of Federal Regulations, Section 50.54(a).
Date of issuance: June 16, 1999.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 152.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 26, 1999 (64
FR 9192).
The April 16, 1999, letter provided clarifying information that did
not change the scope of the original application and expand the initial
proposed no significant hazards consideration determination as
published in the Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: March 9, 1999.
Brief description of amendment: This amendment modifies the
Technical Specifications to increase the inservice inspection interval,
and reduces the scope of volumetric and surface examinations for the
reactor coolant pump flywheels.
Date of issuance: June 8, 1999.
Effective date: June 8, 1999.
Amendment No.: 232.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24196).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: September 30, 1998.
Brief description of amendment: The amendment corrected the
description of the reactor coolant system leakage detection capability
of the reactor building atmosphere gaseous radioactivity monitor in the
Improved Technical Specification Bases and the Final Safety Analysis
Report.
Date of issuance: June 14, 1999.
Effective date: June 14, 1999.
Amendment No.: 179.
Facility Operating License No. DPR-31: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64116).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 14, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: December 3, 1996.
Brief description of amendment: The amendment incorporates certain
improvements from the Standard Technical Specifications for Babcock and
Wilcox plants (NUREG-1430).
Date of issuance: June 15, 1999.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 211.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66708).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 15, 1999.
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.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: January 22, 1999.
Brief description of amendment: Revises Technical Specification
(TS) Section 4.3, ``Fuel Storage,'' by updating the criticality
requirements (k-infinity and U-235 enrichment limits) for storage of
fuel assemblies in the spent fuel racks. This change would allow for
storage of nuclear fuel assemblies with new designs, including GE-12
with a 10X10 pin array.
Date of issuance: June 8, 1999.
Effective date: June 8, 1999.
Amendment No.: 226.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9192).
[[Page 35219]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: October 15, 1998, as
supplemented on December 21, 1998.
Brief description of amendment: Revise the Technical Specifications
(TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment
System Instrument Air System,'' and revises (TS) 3.6.1.3, ``Primary
Containment Isolation Valves,'' Condition E.
Date of issuance: June 9, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 227.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999
(64FR9193).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: April 19, 1999.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.8.1.2, ``Electrical Power Systems, Shutdown,''
and its associated bases to provide a one-time extension of the 18-
month surveillance interval for specific surveillance requirements
associated with the emergency diesel generators for Units 1 and 2. The
surveillances will be performed prior to the first entry into Mode 4
following the current plant shutdown. In addition, for Unit 2 only, a
minor administrative change is included to delete a reference to TS
4.0.8, which is no longer applicable. For Unit 1 only, an editorial
change is made to add the word ``or'' to action statement 3.8.1.2.
Date of issuance: June 8, 1999.
Effective date: June 8, 1999, with full implementation within 45
days.
Amendment Nos.: 228 and 211.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 29, 1999 (64 FR
23129).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: May 15, 1998, as supplemented by
letters dated September 25, October 13, December 9 (two letters), 1998;
January 11, April 1, and April 22, 1999.
Brief description of amendment: This amendment changes Technical
Specification (TS) 5.5, ``Storage of Unirradiated and Spent Fuel,'' to
reflect a planned modification to increase the storage capacity of the
spent fuel pool from 2776 to 4086 fuel assemblies. It also deletes an
inappropriate statement and reference within TS 5.5.
Date of issuance: June 17, 1999.
Effective date: This license amendment is effective as of the date
of its issuance to be implemented before spent fuel is stored within
the new high-density spent fuel rack modules authorized for
installation and use by this amendment.
Amendment No.: 167.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1998 (63
FR 64973).
The September 25, October 13, December 9 (two letters) 1998,
January 11, April 1, and April 22, 1999, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Southern Nuclear Operating Company, Inc., 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 application for amendments: January 21, 1999, which
superseded application dated July 22, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications high radiation trip setpoints for the reactor
building and the refueling floor ventilation exhaust monitors.
Date of issuance: June 9, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--216; Unit 2--157.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24200); this supersedes the original notice dated August 26, 1998 (63
FR 45529).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 30, 1999.
Brief description of amendments: The amendments deleted Technical
Specification 3/4.3.3.4, ``Meteorological Instrumentation,'' and its
associated Bases. These requirements have already been relocated to the
Technical Requirements Manual (TRM). Because the TRM is incorporated
within the South Texas Project updated final safety analysis report for
the units, changes to the relocated requirements will be controlled by
10 CFR 50.59.
Date of issuance: June 16, 1999.
Effective date: June 16, 1999, to be implemented within 30 days.
Amendment Nos.: Unit 1--111; Unit 2--98.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24201).
[[Page 35220]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 16, 1999.
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, Texas 77488.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: February 16, 1999.
Brief Description of amendments: The amendments revise Technical
Specifications (TS) Sections 3.6, 3.9, and 3.16 and the associated
Bases for those sections for Units 1 and 2. The changes consolidate the
auxiliary feedwater cross-connect requirements by relocating the
electrical power requirements from Section 3.16 to Section 3.6. The TS
are also clarified with regard to permitting simultaneous entry into
certain conditions of operation on Units 1 and 2.
Date of issuance: June 7, 1999.
Effective date: June 7, 1999.
Amendment Nos.: 220 and 220.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24203).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 7, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Notice of Issuance of Amendment to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By July 30, 1999, 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
[[Page 35221]]
results of the proceeding. The petition should specifically explain the
reasons why intervention should be permitted with particular reference
to the following factors: (1) the nature of the petitioner's right
under the Act to be made a party to the proceeding; (2) the nature and
extent of the petitioner's property, financial, or other interest in
the proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 10, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications TS 3.7.9, ``Control Room Area Ventilation
System (CRAVS),'' to establish actions to be taken for an inoperable
control room ventilation system due to a degraded control room pressure
boundary. This revision approves a one-time-only action for two CRAVS
trains inoperable due to a degraded control room boundary in Modes 1,
2, 3, and 4, that is to be completed within 24 hours. The applicable TS
Bases have been revised to document the TS changes and to provide
supporting information.
Date of issuance: June 11, 1999.
Effective date: As of the date of issuance and shall be implemented
upon receipt.
Amendment Nos.: Unit 1--185; Unit 2--167.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
The Commission's related evaluation and the amendment, finding of
emergency circumstances, consultation with the State of North Carolina,
and final no significant hazards consideration determination are
contained in a Safety Evaluation dated June 11, 1999.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Dated at Rockville, Maryland, this 23rd day of June 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-16489 Filed 6-29-99; 8:45 am]
BILLING CODE 7590-01-P