[Federal Register Volume 64, Number 221 (Wednesday, November 17, 1999)]
[Notices]
[Pages 62704-62722]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-29846]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from October 23, 1999, through November 5, 1999.
The last biweekly notice was published on November 3, 1999 (64 FR
59796).
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 December 17, 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 electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). 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 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
[[Page 62705]]
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: 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 electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
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: October 21, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for the Harris Nuclear Plant
(HNP) to implement selected improvements described in NRC Generic
Letter (GL) 93-05, ``Line-Item Technical Specifications To Reduce
Surveillance Requirements For Testing During Power Operation,'' dated
September 27, 1993. Specifically, HNP proposes to modify the following
TS to be consistent with GL 93-05: (1) TS 4.1.3.1.2--Change the
frequency of the control rod movement test to quarterly; (2) TS
4.6.4.1--Change the frequency of the Hydrogen Monitor analog channel
operational test to quarterly; (3) TS 4.3.3.1 (Table 4.3-3)--Change the
Radiation Digital Channel Operational Test to quarterly; (4) TS
4.4.6.2.2.b.--Change the time for remaining in cold shutdown without
leak testing the Reactor Coolant System Pressure Isolation Valves to 7
days; (5) TS 4.4.3.2--Change the testing of the capacity of pressurizer
heaters to once per 18 months; (6) TS 4.6.4.2.a.--Change the Hydrogen
Recombiner functional test to once per 18 months; and (7) TS
4.7.1.2.1.a--Change frequency of testing Auxiliary Feedwater Pumps to
quarterly.
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.
There are no systems being modified as a result of this change.
Additionally, the way in which equipment is tested is not affected
by this change. Reducing surveillance intervals for TS components
(such as control rod testing) may reduce the probability of an
accident (rod drop accident) by reducing actions that could cause an
accident to occur (rod movement).
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.
No system, structure, or component is being modified as a result
of this change. Additionally, there are no changes to the way
equipment is operated as a result of this change. Operating
parameters are not being modified as a result of this change.
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.
These proposed changes are in accordance with NRC Generic Letter
93-05, dated September 27, 1993 and NUREG-1366, dated December 1992.
These changes pertain to testing requirements for TS equipment which
help ensure operability requirements are met. This change does not
modify the required safety function or operating parameters for
equipment described in HNP TS.
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.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Kahtan Jabbour, Acting.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: October 15, 1999.
[[Page 62706]]
Description of amendment request: The amendments would revise
Section 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of
the Technical Specifications. Section 5.5.7 currently specifies that
inspections be done according to Regulatory Position c.4.b of
Regulatory Guide 1.14, Revision 1, such that an in-place ultrasonic
volumetric examination of the areas of higher stress concentration at
the bore and keyway be performed at approximately 3-year intervals. The
licensee proposed to revise this to require a qualified in-place
ultrasonic examination over the volume from the inner bore of the
flywheel to the circle of one half the outer radius, or a surface
examination (magnetic particle and/or penetration testing) of exposed
surfaces defined by the volume of the disassembled flywheel. The
licensee stated that the technical basis has been set forth in
Westinghouse Topical Report WCAP-14535A, and cited similar amendments
already granted to other nuclear plants.
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:
First Standard
Would implementation of the changes proposed in this LAR involve
a significant increase in the probability or consequences of an
accident previously evaluated?
No. There are no accident probabilities or consequences impacted
by this LAR [license amendment request]. As discussed in Attachment
3 [the licensee's description of the proposed amendment], following
a reduction in the scope and frequency of the examinations currently
required by the applicable Technical Specifications and Regulatory
Guide 1.14, Revision I, an adequate inservice inspection program
will continue to be maintained for the reactor coolant pump
flywheels. Since the integrity of the flywheels will continue to be
ensured, these components will continue to be available to fulfill
their existing design function during pump coastdown flow
transients. Additionally, there is no more risk that the flywheels
will become a source of missile generation. Consequently, there is
no significant increase in the probability or consequences of an
accident previously evaluated.
Second Standard
Would implementation of the changes proposed in this LAR create
the possibility of a new or different kind of accident from any
previously evaluated?
No. The proposed changes contained in this LAR only reduce the
existing inspection requirements for the reactor coolant pump
flywheels. This LAR proposes no changes to the plants' design,
equipment, or method of operation at either McGuire or Catawba
Nuclear Station. Furthermore, the reduction in the inspection
requirements for the flywheels has been generically approved by the
NRC and is justified by WCAP-14535A. Therefore, since implementation
of this LAR results in no actual impact upon either of the Duke
nuclear plants, and since the integrity of the flywheels will
continue to be ensured at an acceptable level, no new or different
kinds of accidents are being created.
Third Standard
Would implementation of the changes proposed in this LAR involve
a significant reduction in a margin of safety?
No. Margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. These barriers are unaffected by the changes proposed in
this LAR. As discussed in WCAP-14535A, a reduction in the frequency
for performing the inservice inspections currently done in
accordance with Regulatory Guide 1.14, Revision I, will not preclude
the ability to accurately demonstrate the integrity of the reactor
coolant pump flywheels. This LAR creates no additional threat to the
integrity of the fission product barriers from the standpoint of
missile generation or otherwise. Therefore, implementation of the
changes proposed in this LAR does not impact the assumption of the
integrity of the flywheels, the fission product barriers, or any
other accident analyses assumptions. Consequently, no margin of
safety will be significantly impacted by this LAR.
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.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
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: October 15, 1999.
Description of amendment request: The amendments would revise
Section 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' of
the Technical Specifications. Section 5.5.7 currently specifies that
inspections be done according to Regulatory Position c.4.b of
Regulatory Guide 1.14, Revision 1, such that an in-place ultrasonic
volumetric examination of the areas of higher stress concentration at
the bore and keyway be performed at approximately 3-year intervals. The
licensee proposed to revise this to require a qualified in-place
ultrasonic examination over the volume from the inner bore of the
flywheel to the circle of one half the outer radius, or a surface
examination (magnetic particle and/or penetration testing) of exposed
surfaces defined by the volume of the disassembled flywheel. The
licensee stated that the technical basis has been set forth in
Westinghouse Topical Report WCAP-14535A, and cited similar amendments
already granted to other nuclear plants.
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:
First Standard
Would implementation of the changes proposed in this LAR involve
a significant increase in the probability or consequences of an
accident previously evaluated?
No. There are no accident probabilities or consequences impacted
by this LAR [license amendment request]. As discussed in Attachment
3 [the licensee's description of the proposed amendment], following
a reduction in the scope and frequency of the examinations currently
required by the applicable Technical Specifications and Regulatory
Guide 1.14, Revision I, an adequate inservice inspection program
will continue to be maintained for the reactor coolant pump
flywheels. Since the integrity of the flywheels will continue to be
ensured, these components will continue to be available to fulfill
their existing design function during pump coastdown flow
transients. Additionally, there is no more risk that the flywheels
will become a source of missile generation. Consequently, there is
no significant increase in the probability or consequences of an
accident previously evaluated.
Second Standard
Would implementation of the changes proposed in this LAR create
the possibility of a new or different kind of accident from any
previously evaluated?
No. The proposed changes contained in this LAR only reduce the
existing inspection requirements for the reactor coolant pump
flywheels. This LAR proposes no changes to the plants' design,
equipment, or method of operation at either McGuire or Catawba
Nuclear Station. Furthermore, the reduction in the inspection
requirements for the flywheels has been generically approved by the
NRC and is justified by WCAP-14535A. Therefore, since implementation
of this LAR results in no actual impact upon either of the Duke
nuclear plants, and since the integrity of the flywheels will
continue to be ensured at an acceptable level, no new or different
kinds of accidents are being created.
[[Page 62707]]
Third Standard
Would implementation of the changes proposed in this LAR involve
a significant reduction in a margin of safety?
No. Margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. These barriers are unaffected by the changes proposed in
this LAR. As discussed in WCAP-14535A, a reduction in the frequency
for performing the inservice inspections currently done in
accordance with Regulatory Guide 1.14, Revision I, will not preclude
the ability to accurately demonstrate the integrity of the reactor
coolant pump flywheels. This LAR creates no additional threat to the
integrity of the fission product barriers from the standpoint of
missile generation or otherwise. Therefore, implementation of the
changes proposed in this LAR does not impact the assumption of the
integrity of the flywheels, the fission product barriers, or any
other accident analyses assumptions. Consequently, no margin of
safety will be significantly impacted by this LAR.
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.
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: September 29, 1999.
Description of amendment request: The proposed amendments would
revise the Containment Inservice Inspection (ISI) Program Technical
Specifications (TS) 5.5.2, ``Containment Leakage Testing Program,'' and
TS 5.5.7, ``Pre-Stressed Concrete Containment Tendon Surveillance
Program.'' The proposed amendments would permit the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section
XI, Subsection IWL visual examinations to be performed in lieu of
concrete and post-tensioning system general visual examinations
required by 10 CFR 50, Appendix J and Regulatory Guide 1.163 between
Type A tests. In addition, the amendment would permit general visual
examinations of the concrete and post-tensioning system that can be
performed with a unit in operation to be performed prior to the
beginning of a refueling outage during which a Type A test is
scheduled.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. Implementation of this amendment would not involve a
significant increase in the probability or consequences of an
accident previously evaluated. Approval of this amendment will have
no significant effect on accident probabilities or consequences. The
containment is not an accident initiating system or structure;
therefore, there will be no impact on any accident probabilities by
the approval of this amendment. The containment serves an important
function to mitigate consequences of postulated accidents previously
evaluated and the examination frequencies proposed in this amendment
will not result in a reduction in the capacity of the containment to
meet its intended function. The requested flexibility in scheduling
containment visual examinations has no significant impact on the
validity of the examinations or of containment structural integrity.
Additionally, the change to Technical Specification 5.5.7 and
the planned revision to Selected Licensee Commitment 16.6.2
described in this amendment application reflect the adoption of an
ASME Section XI, Subsection IWE and IWL Inservice Inspection Program
as required by 10 CFR 50 Section 55a(g)(4). Implementation of this
program will not result in a reduction in the capacity of the
containment to meet its intended function.
Therefore, the probability or consequences of an accident
previously evaluated will not be increased by approval of the
requested changes.
B. Create the possibility of a new or different kind of accident
from the accident previously evaluated?
No. Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant that would introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators, since the containment
functions primarily as an accident mitigator.
C. Involve a significant reduction in a margin of safety?
No. Implementation of this amendment would not involve a
significant reduction in a margin of safety. Margin of safety is
related to the confidence in the ability of the fission product
barriers to perform their design functions during and following an
accident situation, including the performance of the containment.
This component is already capable of performing as intended, and its
function is verified by visual examination, post-tensioning system
examinations, and leakage rate testing.
The examination requirements of ASME XI, Subsection IWL, are
essentially identical to those contained in Regulatory Guide 1.35,
Rev. 3, and are more rigorous than those required by 10 CFR 50,
Appendix J and Regulatory Guide 1.163. Previous visual examinations
of containment concrete and post-tensioning system surfaces have not
revealed any indications of abnormal degradation of the containment.
The five-year frequency for IWL examinations is adequate in lieu of
the general visual examination frequency specified in Regulatory
Guide 1.163 for containment concrete and post-tensioning system
examinations.
The ability of the containment to perform its design function
will not be impaired by the implementation of this amendment at
Oconee Nuclear Station. Consequently, no safety margins will be
impacted.
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.
Attorney for licensee: Anne W. Cottingham, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of amendment request: June 17, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3.4.9.1 and associated
figures to extend the applicability of the heatup and cooldown curve
pressure and temperature limits from 10 effective full power years
(EFPY) to 15 EFPY. The proposed changes include new heatup and cooldown
curves developed in accordance with the methodology provided in
Regulatory Guide 1.99, Revision 2, and Code Case N-640. The
applicability of TS Section 3.4.9.3, Overpressure Protection Systems,
is also updated to 15 EFPY, and the maximum allowable power operated
relief valve (PORV) setpoints for the over pressure protection system
are revised. Revisions to the TS Bases are also made.
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 62708]]
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed heatup and cooldown curves have been revised by
changing the applicability from 10 effective full power years (EFPY)
to 15 EFPY. The curves have been developed in accordance with the
methodology provided in Regulatory Guide 1.99, Revision 2 and Code
Case N-640. The proposed heatup and cooldown curves define limits
that still ensure the prevention of nonductile failure for the
reactor vessel. The design basis events that were protected against
have not changed; therefore, the probability of an accident is not
increased.
The overpressure protection system (OPPS) has been revised such
that the applicability has changed from 10 EFPY to 15 EFPY. This
system protects the Reactor Coolant System (RCS) at low temperatures
so that the integrity of the Reactor Coolant Pressure Boundary
(RCPB) is not compromised by violating the pressure/temperature (P/
T) limits. These changes were determined in accordance with the
methodologies set forth in the regulations to provide an adequate
margin of safety to ensure the reactor vessel will withstand the
effects of normal cyclic loads due to temperature and pressure
changes as well as the loads associated with postulated faulted
events. The lower limit on pressure during the design basis OPPS
mass injection and heat addition transients is established based on
operational consideration for the RCP number one seal limit which
requires a nominal differential pressure across the seal faces for
proper film-riding performance. As part of the OPPS setpoint
evaluation, margin to the RCP number one seal limit is evaluated.
This limit corresponds to a differential pressure across the
seal of 200 psid, which corresponds to the gage pressures. The
pressure undershoot below the PORV setpoint during a design basis
mass injection or heat addition event can exceed 100 psi. Therefore,
with the PORV setpoints developed for the 15 EFPY heatup and
cooldown curves, there is the potential for RCS pressure to violate
the RCP number one seal limit at the lowest RCS temperatures.
Undershoot below the PORV setpoint can be significantly higher
if both PORVs actuate during an OPPS event, and it is anticipated
that the pump seal limit would be exceeded. However, staggering the
setpoints minimizes the likelihood that both PORVs will actuate
simultaneously during credible OPPS events. Similarly, WCAP 14040-
NP-A indicates that when there is insufficient range between the
upper and lower pressure limits to select PORV setpoints that
provide protection against violating both limits, then the setpoint
selection that provides protection against the upper limit violation
takes precedence. WCAP-4040-NP, Revision 1 was approved by the NRC
by letter dated October 16, 1995, which was incorporated in Revision
2 of the approved WCAP issued in January 1996.
Modification of the heatup and cooldown curves and OPPS
setpoints does not alter any assumptions previously made in the
radiological consequence evaluations nor affect mitigation of the
radiological consequences of an accident described in the Updated
Final Safety Analysis Report (UFSAR). Therefore, the proposed
changes will not significantly increase 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 heatup and cooldown curves applicable for the first
15 EFPY were generated using approved methodology and Code Case N-
640. Generating these curves with Code Case N-640 reduced the excess
conservatism that exists in the current curves and results in an
increase in the safety of the plant, as the likelihood of RCP seal
failures and/or fuel problems will decrease. The change does not
cause the initiation of any accident nor create any new single
failure.
The modification of the OPPS setpoints ensures that the RCPB
integrity is protected at low temperatures. The new setpoints were
selected using conservative assumptions to ensure that sufficient
margin is available to prevent violation of the P/T limits due to
anticipated mass and heat input transients. The modification of the
setpoints does not change, degrade, or prevent the safe response of
the RCS to accident scenarios, as described in UFSAR Chapter 15. The
proposed change does not cause the initiation of any accident nor
create any new credible single failure.
Therefore, the proposed license amendment does 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 new P/T curves define the limits for ensuring prevention of
nonductile failure for the reactor vessel, and does not
significantly reduce the margin of safety for the plant. The
methodology provided in Code Case N-640 removed some of the excess
conservatism from the current Appendix G analysis. However, this
improved overall plant safety by expanding the operating window
relative to the RCP seal requirements. The probability of damaging
the RCP seals is reduced. Therefore, the margin of safety is not
significantly reduced.
The OPPS setpoints will continue to ensure the RCS pressure
boundary will be protected from pressure transients. They were
generated using the proposed heatup and cooldown curves as input.
The OPPS setpoints include additional margin by including instrument
uncertainties not included in the current setpoints. Therefore, the
margin of safety is not significantly 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.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Sheri R. Peterson.
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: July 15, 1999.
Description of amendment request: The license amendment request
(LAR) proposes to revise the Technical Specifications frequency for the
Quench and Recirculation Spray Systems nozzle air flow test from 5
years to 10 years. This LAR also includes a revision to correct the
terminology used in an action requirement as well as miscellaneous
editorial and format 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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed extension of the testing frequency of the Quench
Spray and Recirculation Spray Systems' nozzles to ten years does not
change the way these systems are operated or their operability
requirements. The proposed change to the surveillance frequency of
safety equipment has no impact on the probability of an accident
occurrence nor can it create a new or different type of accident.
NUREG-1366, ``Improvements to Technical Specifications Surveillance
Requirements,'' dated December 1992, and Generic Letter 93-05,
``Line Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,''
dated September 27, 1993, concluded that the corrosion of stainless
steel piping is negligible during the extended surveillance interval
for nozzle testing. The results of the above NRC study were
evaluated by Duquesne Light Company and found to be applicable to
Beaver Valley Power Station (BVPS) Unit 1 and 2. Since the Quench
Spray and Recirculation Spray Systems are maintained dry, there is
no additional mechanism that could cause blockage of the spray
nozzles. Thus, the nozzles in these spray systems are expected to
remain operable during the ten year surveillance interval to
mitigate the consequence of an accident previously evaluated. No
obstructed or clogged spray systems' nozzles have been observed
during the five year frequency surveillance tests at either BVPS
Unit 1 or Unit 2 to date. Testing of the spray systems' nozzles at
the proposed reduced frequency will not increase the probability of
occurrence of a postulated accident or the consequences of an
accident previously evaluated.
This license amendment also revises the Action criteria in the
BVPS Unit 1 and 2 Axial Flux Difference [AFD] technical
[[Page 62709]]
specification to correct the terminology referring to the Core
Operating Limits Report (COLR) limits. The proposed change
incorporates the terminology (acceptable operation limits) used in
the corresponding Action condition of the ISTS [Improved Standard
Technical Specifications]. The proposed change does not alter the
AFD limits specified in the COLR and the AFD specification continues
to assure plant operation within those limits. With AFD within the
acceptable operation limits specified in the COLR, the resulting
axial power distribution remains within the initial conditions
assumed in the safety analyses. Therefore, these changes will not
increase the probability of occurrence of a postulated accident or
the 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 reduced frequency testing of the Quench Spray and
Recirculation Spray Systems' nozzles does not change the way the
spray systems are operated. The reduced frequency of testing the
spray nozzles does not change the plant operation or system
readiness. The reduced frequency testing of the Quench Spray and
Recirculation Spray Systems' nozzles does not generate any new
accident precursors. Therefore, the possibility of a new or
different kind of accident previously evaluated is not created by
the proposed changes in surveillance frequency of the spray systems'
nozzles.
This license amendment also revises the Action criteria in the
BVPS Unit 1 and 2 Axial Flux Difference technical specification to
correct the terminology referring to the Core Operating Limits
Report (COLR) limits. This addresses an incorrect use of terminology
and the revision does not involve a technical intent change.
Therefore, the possibility of a new or different kind of accident
previously evaluated is not created by the proposed terminology
correction.
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 the mitigation and
control of design bases 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.
Reduced testing of the Quench Spray and Recirculation Spray
Systems' nozzles does not change the way these spray systems are
operated or these spray systems' operability requirements. Generic
Letter 93-05 and NUREG-1366 concluded that the corrosion of
stainless steel piping is negligible during the extended
surveillance interval for nozzle testing. The results of the above
NRC study were evaluated by Duquesne Light Company and found to be
applicable to BVPS Unit 1 and 2. Since the Quench Spray and
Recirculation Spray Systems are maintained dry, there is no
additional mechanism that could cause blockage of these spray
systems' nozzles. Thus, the proposed reduced testing frequency is
adequate to ensure spray nozzle operability. The surveillance
requirements do not affect the margin of safety in that the
operability requirements of the Quench Spray and Recirculation Spray
Systems remain unaltered. The existing safety analyses remain
bounding. Therefore, the margin of safety is not adversely affected.
This license amendment also revises the Action criteria in the
BVPS Unit 1 and 2 Axial Flux Difference technical specification to
correct the terminology referring to the Core Operating Limits
Report (COLR) limits. This addresses an incorrect use of terminology
and the revision does not involve a technical intent change. The
operating criteria on Axial Flux Difference are not altered from
their intended requirements. Therefore, the margin of safety is not
adversely affected by the proposed terminology correction.
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.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Sheri R. Peterson
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: July 20, 1999
Description of amendment request: The licensee amendment request
proposes to relocate the following Technical Specifications items to
the Licensing Requirements Manual:
In-core Detectors (Unit 1 and 2),
Chlorine Detection System (Unit 1 and 2),
Turbine Over-speed Protection (Unit 2 only),
Crane Travel Spent Fuel Storage Pool Building (Unit 1 and 2).
In addition to the relocation, certain editorial and format changes
are proposed. Also, it is proposed that certain information on the
Remote Shutdown Panel Monitoring Instrumentation be moved to the
Updated Final Safety Analysis Report (USFAR).
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?
Consistent with the guidance provided in Generic Letter (GL) 95-
10 and the content of the Improved Standard Technical Specifications
(ISTS) contained in NUREG-1431, Rev. 1, this license amendment
request (LAR) proposes the relocation of the following TS to the
Licensing Requirements Manual (LRM):
3/4.3.3.2 Incore Detectors (Unit 1 and 2)
3/4.3.3.7 Chlorine Detection System (Unit 1 and 2)
3/4.3.4 Turbine Overspeed Protection (Unit 2 only)
In order to completely relocate the chlorine detection system
requirements from the Technical Specifications (TS), portions of the
Unit 1 Specifications 3/4.7.7, Control Room Habitability Systems and
3/4.9.15, Control Room Emergency Habitability Systems, as well as
the Unit 2 Specification, 3/4.7.7, Control Room Emergency Air
Cleanup and Pressurization System are proposed to be revised to
reflect the removal of the chlorine detection system from the TS.
The applicable surveillance requirements, and modes of applicability
from these specifications are proposed to be relocated to the LRM
along with the associated chlorine detection system TS. In addition,
new actions have been added to the chlorine detection system
specifications to integrate the new requirements.
In addition to the TS identified for relocation by the NRC in GL
95-10, this LAR proposes the relocation of another TS that does not
meet the criteria of 10 CFR 50.36 and is not included in the ISTS.
The additional TS proposed to be relocated to the LRM is 3/4.9.7
Crane Travel Spent Fuel Storage Pool Building (Unit 1 and 2).
This LAR also proposes that the TS Bases section associated with
each of the TS listed above be relocated to the LRM as well. The
appropriate TS pages (i.e., LCO, Bases, Table of Contents, etc.) are
revised to reflect the removal of these Specifications and Bases
from the TS.
The TS and bases discussed above and proposed for relocation
will be moved into the BVPS LRM. The Unit 1 and Unit 2 LRM are
appendices of the associated unit UFSAR. As part of the UFSAR any
changes made to the LRM must be in accordance with the provisions of
10 CFR 50.59.
In addition to the relocation of the above listed TS, this LAR
includes the removal of the ``Measurement Range'' information from
the Unit 1 and 2 TS Table 3.3-9, Remote Shutdown Panel Monitoring
Instrumentation. This design information is being moved from the TS
to an applicable Updated Final Safety Analysis Report (UFSAR)
section. The removal of this detail from the TS is consistent with
the level of detail in the corresponding ISTS Specification. As part
of the UFSAR any changes made to the measurement range information
must be in accordance with the provisions of 10 CFR 50.59.
LAR 1A-251/2A-121 includes two Bases enhancements. Additional
information is being added to the reactor trip system
instrumentation Bases to discuss diverse and anticipatory protection
features not credited in the accident analyses. The reactor trip
system instrumentation Bases is also revised
[[Page 62710]]
to more clearly describe the source and intermediate range neutron
flux protection features required during shutdown modes.
The proposed changes include the addition of license numbers to
some of the TS pages contained in this LAR. In addition, this LAR
contains changes that update the format of the affected TS pages and
make editorial corrections. These changes are administrative in
nature and do not impact the technical content of the affected TS
pages.
The proposed changes regarding the relocation of information
from the TS in this LAR follow the guidance provided in Generic
Letter 95-10, the NRC ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors'' (58 FR
39132) dated July 22, 1993, and are consistent with the content of
the ISTS. In addition, the proposed location for this information
(UFSAR and LRM) ensures that future changes to the relocated
requirements will be in accordance with the provisions of 10 CFR
50.59 and that NRC review and approval will be requested should a
change to this information involve an unreviewed safety question.
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 conditions and
assumptions for accidents described in the UFSAR remain as
previously analyzed. The failure of any of the systems or components
affected by this LAR, except for turbine overspeed protection, is
not an accident initiating event. Due to the low likelihood of
equipment damage or failure resulting from turbine missiles
generated by a turbine overspeed event, assumptions related to the
turbine overspeed protection system are not part of an initial
condition of a design basis accident or transient.
The proposed amendment also does not involve a significant
increase in the consequences of an accident previously evaluated.
The amendment does not reduce the current requirements for the
systems and components proposed for relocation. The amendment only
requests that the requirements be retained in a more appropriate
document. The systems and components proposed for relocation in this
amendment perform no active role in mitigating a design basis
accident described in the UFSAR. The systems or components proposed
for relocation are not part of the initial conditions assumed in a
safety analysis for a design basis accident described in the UFSAR.
In addition, the affected systems and components do not function to
actuate any protective equipment, nor are they part of the primary
success path assumed in the safety analyses to mitigate any design
basis accident described in the UFSAR.
The bases enhancements included in this LAR are administrative
in nature and serve only to provide additional descriptive
information. These changes do not impact 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 any plant systems.
Moving specifications to the LRM or design information to the
UFSAR will not change the physical plant or the modes of plant
operation. Whether these specifications are located in the TS or the
LRM has no effect on any previously evaluated accident. The
relocation of TS information does not involve a change in the
configuration of equipment nor does it alter the design or operation
of plant systems.
Expanding the Bases for both units to discuss additional
information regarding the protective functions not credited in the
safety analysis or the neutron flux trip functions required in
shutdown modes provides additional information to enhance the
awareness of the protective instrumentation functions. The proposed
bases changes do not result in any adjustments or physical
alteration to the affected protective instrumentation functions. The
Reactor Protection System will continue to function as currently
designed and assumed in the accident analyses.
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 margin of safety depends on the maintenance of specific
operating parameters and systems within design requirements and
safety analysis assumptions.
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 the mitigation and
control of design bases 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, and
sufficient instrumentation and control capability is available for
monitoring and maintaining the unit status.
The relocation of TS requirements and information to the LRM or
UFSAR does not reduce the requirements for the affected systems and
components to be maintained operable and function within design
requirements. The relocation of TS requirements and information to
the LRM and UFSAR will allow changes to this information to be made
in accordance with the provisions of 10 CFR 50.59 and continues to
ensure that NRC review and approval will be requested should a
change to this information involve an unreviewed safety question.
Expanding the Bases for both units to discuss additional
information regarding the protective functions not credited in the
safety analysis or the neutron flux trip functions required in
shutdown modes provides additional information to enhance the
awareness of the protective instrumentation functions. The addition
of descriptive text to the TS bases does not affect the TS
requirements for the affected equipment to be maintained operable
and function within the applicable design requirements. The Reactor
Protection System will continue to function as currently designed
and assumed in the accident analyses.
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.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Sheri R. Peterson.
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: September 20, 1999.
Description of amendment request: The proposed amendments would
revise the standard to which the control room ventilation charcoal and
Supplementary Leak Collection and Release System (SLCRS) charcoal must
be laboratory tested as specified in: Beaver Valley Power Station, Unit
No. 1 (BVPS-1), Technical Specification (TS) 4.7.7.1.1.c.2 for the
Control Room Emergency Habitability Systems; BVPS-1 TS 4.7.8.1.b.3 for
the SLCRS; Beaver Valley Power Station, Unit No. 2 (BVPS-2), TS
4.7.7.1.d for the Control Room Emergency Air Cleanup and Pressurization
System; and BVPS-2 TS 4.7.8.1.b.3 for the SLCRS. NRC Generic Letter 99-
02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated
June 3, 1999, requested licensees to revise their TS criteria
associated with laboratory testing of ventilation charcoal to a valid
test protocol, which included American Society for Testing Materials
(ASTM) D3803-1989. This license amendment request revises the charcoal
laboratory standard to follow ASTM D3803-1989 for each BVPS Unit.
This license amendment request also: (1) Revises the minimum amount
of output in kilowatts needed for the control room emergency
ventilation system heaters at each BVPS Unit; (2)
[[Page 62711]]
revises BVPS-1 SLCRS surveillance testing criteria to be consistent
with American National Standards Institute/American Society of
Mechanical Engineers (ANSI/ASME) N510-1980, the BVPS-1 control room
ventilation testing, and the BVPS-2 SLCRS/control room ventilation
testing; and (3) makes minor typographical corrections and editorial
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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to the surveillance requirements for the
laboratory testing of ventilation system charcoal are consistent
with Generic Letter 99-02. The proposed change will adopt ASTM
D3803-1989 as the laboratory testing standard for performing the
surveillance associated with the Control Room emergency ventilation
and the SLCRS charcoal filters at each BVPS Unit. Thus this proposed
change will not involve a significant increase in the probability or
consequences of a previously evaluated accident since this standard
provides the assurance for continuing to comply with the current
BVPS Unit 1 and Unit 2 licensing basis as it relates to the dose
limits of GDC 19 and 10 CFR Part 100.
The change in the control room emergency ventilation system
heater minimum output at both BVPS Units does not change the system
ability to meet its design bases. The change in the BVPS Unit 1
SLCRS testing frequency for adsorber/filter in-place testing and the
adsorber laboratory testing does not change the SLCRS system's
ability to meet its design bases. The change in the BVPS Unit 1
SLCRS testing frequency for SLCRS air flow distribution testing does
not change the SLCRS system's ability to meet its design bases.
Therefore, these changes will not increase the probability of
occurrence of a postulated accident or the consequences of an
accident previously evaluated since these systems' ability to
operate as required remains unchanged.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed license amendment to the control room emergency
ventilation system and SLCRS at both BVPS Units does not change the
way the system is operated. The proposed changes only involve
changes to the surveillance testing. These testing modifications do
not alter these systems' ability to perform their design bases.
Therefore, these proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated
accident since the control room emergency ventilation system and
SLCRS will continue to operate in accordance with their previous
design bases.
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
system, structures or components important to the mitigation and
control of design bases 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.
The proposed license amendment to the control room emergency
ventilation system and SLCRS at both BVPS Units does not change the
way the system is operated. The proposed changes only involve
changes to the surveillance testing. These testing modifications do
not alter these systems' ability to perform their design bases. The
existing safety analyses remain bounding. Therefore, the margin of
safety is not adversely affected.
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.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Sheri R. Peterson.
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit 2, Shippingport, Pennsylvania
Date of amendment request: September 22, 1999.
Description of amendment request: The proposed amendment would
allow a one-time only extension to the surveillance interval of
Technical Specification Surveillance 4.7.12.d for functional testing of
snubbers. The proposed extension would be limited to the end of the 8th
refueling outage or November 30, 2000, whichever occurs sooner.
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 is for a one-time extension to the
surveillance interval for functional testing of snubbers specified
in Technical Specification (TS) 4.7.12.d. The proposed change
involves revising the calendar time allowed between functional tests
and would result in a maximum surveillance interval extension of
approximately 6.5 months.
The proposed change continues to adequately limit plant
operation between required snubber surveillances by ensuring the
required surveillances are performed by November 30, 2000.
Therefore, the proposed change continues to limit snubber wear due
to vibration and elevated temperatures. The elevated temperatures
and vibration experienced during plant operation are the primary
contributors to snubber wear.
In addition, snubber-testing experience has shown that the
historical failure rate of snubbers is low. There have been seven
refueling outages since Unit 2's startup in 1987. Only during the
first refueling outage, 2R01, did the snubber functional test sample
plan identify any inoperable snubbers. In that outage, seven
snubbers tested inoperable. All failed due to damage sustained
during original construction and startup activities. Since 2R01, no
inoperable snubbers were found by sample plan functional testing
performed during each surveillance interval. Also, the latest visual
inspections performed on the Unit 2 snubbers (during 2R07) revealed
no evidence of damage or potential problems with any snubber.
Due to the low incidence of snubber functional test failures
resulting from sample plan testing and the limited plant operating
time between tests, the possibility of a snubber failure resulting
from this one-time surveillance extension is low. No changes are
being made to any accident initiator. No analyzed accident scenario
is being changed. The initiating conditions and assumptions of
previously analyzed accidents remain unchanged. Therefore, the
proposed change does not involve a significant increase in the
probability of a previously evaluated accident.
This change does not involve a physical change to the plant and
does not affect the acceptance criteria specified in the TS for
snubber functional testing, nor does this change reduce the remedial
actions required for inoperable snubbers. Therefore, the proposed
change does not involve a significant increase in the 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 any plant systems. The one-time surveillance
interval extension proposed by this change will not reduce the
capability of the snubbers to perform their design function.
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 margin of safety depends on the maintenance of specific
operating parameters
[[Page 62712]]
and systems within design requirements and safety analysis
assumptions.
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 the mitigation and
control of design bases 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, and
sufficient instrumentation and control capability is available for
monitoring and maintaining the unit status.
The proposed change is for a one-time extension to the
surveillance interval for functional testing of snubbers specified
in Technical Specification 4.7.12.d. The proposed change continues
to adequately limit plant operation between required snubber
surveillances by ensuring the required surveillances are performed
by November 30, 2000. Therefore, the proposed change continues to
limit snubber wear due to vibration and elevated temperatures. The
elevated temperatures and vibration experienced during plant
operation are the primary contributors to snubber wear.
In addition, snubber-testing experience has shown that the
historical failure rate of snubbers is low. There have been seven
refueling outages since Unit 2's startup in 1987. Only during the
first refueling outage, 2R01, did the snubber functional test sample
plan identify any inoperable snubbers. In that outage, seven
snubbers tested inoperable. All failed due to damage sustained
during original construction and startup activities. Since 2R01, no
inoperable snubbers were found by sample plan functional testing
performed during each surveillance interval. Also, the latest visual
inspections performed on the Unit 2 snubbers (during 2R07) revealed
no evidence of damage or potential problems with any snubber.
This change does not involve a physical change to the plant and
does not affect the acceptance criteria specified in the TS for
snubber functional testing, nor does this change reduce the remedial
actions required for inoperable snubbers. The snubbers and systems
supported by the snubbers will continue to be available to perform
their intended safety functions during the requested extension
period.
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.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Sheri R. Peterson.
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: July 30, 1999.
Description of amendment request: The request proposes changes to
the Technical Specifications (TSs) and the operating license to extend
operation of the station from its licensed power of 2894 megawatts
thermal (MWt) to the uprated power level of 3039 MWt, an increase of 5
percent. The proposed changes are to (1) extend the definition of rated
thermal power in TS Section 1.1 and the operating license to 3039 MWt;
(2) reduce the thermal power safety limit of TSs 1.4, 2.1.1.1, 3.2.1,
3.2.2, 3.2.3, 3.3.1.1, 3.4.3, and 3.7.5; (3) increase the reactor steam
dome pressure in TS Table 3.1.4-1, TS 3.4.12, and SR 3.5.3.3; (4)
increase the control rod drive charging water header pressure in TSs
3.1.5, 3.9.5, and 3.10.8; (5) increase the standby liquid control (SLC)
system Boron-10 enrichment and concentration criteria in TS 3.1.7; (6)
increase the surveillance test discharge pressure for the SLC pump in
surveillance requirement (SR) 3.1.7.7; (7) increase the allowable value
of the reactor vessel steam dome pressure--high scram setpoint in TS
Table 3.3.1.1-1; (8) increase the allowable value for the anticipated
transient without scram--reactor pressure trip reactor steam dome
pressure--high setpoint in SR 3.3.4.2.4; (9) revise the safety, relief,
and low low set function of the main steam safety/relief valves (SRVs)
in SRs 3.3.6.4.3 and 3.4.4.1; (10) increase the upper and lower bounds
on reactor pressure for the purposes of performing reactor core
isolation cooling pump flow rate surveillance at high pressure in SR
3.5.3.3; (11) increase the main steam line flow--high reactor isolation
trip in TS Table 3.3.6.1-1; (12) reduce the thermal power limits for
single loop operation in TS 3.4.1; (13) increase the upper and lower
bounds on reactor pressure for the purposes of performing pressure
isolation valve surveillance at high pressure in SR 3.4.6.1; and (14)
revise the reactor coolant system pressure/temperature limits in TS
3.4.11 (including replacing TS Figure 3.4.11-1 with figures for 14 and
32 effective full power years of operation). Item (9) includes
increasing the main steam SRV setpoint tolerance from +0%, -2% to [plus
or minus] 3% in SR 3.4.4.1.
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) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The increase in power level discussed herein will not
significantly increase the probability or consequences of an
accident previously evaluated.
The probability (frequency of occurrence) of Design Basis
Accidents occurring is not affected by the increased power level, as
the regulatory criteria established for plant equipment (ASME
[American Society of Mechanical Engineers] Code, IEEE [Institute of
Electrical and Electronic Engineers] standards, NEMA [National
Equipment Manufacturers Association] standards, Reg[ulatory] Guide
criteria, etc.) will still be complied with at the uprated power
level. An evaluation of the BWR [boiling water reactor]
probabilistic risk assessments concludes that the calculated core
damage frequencies will not significantly change due to [the] power
uprate. Scram setpoints (equipment settings that initiate automatic
plant shutdowns) will be established such that there is no
significant increase in scram frequency due to [the] uprate. No new
challenges to safety-related equipment will result from [the] power
uprate.
The changes in consequences of hypothetical accidents which
would occur from 102% of the uprated power, compared to those
previously evaluated from [greater than or equal to] 102% of the
original power, are in all cases insignificant, because the accident
evaluations from [the] power uprate to 105% of original power
([approximately] 106% of original steam) flow will not result in
exceeding the NRC-approved acceptance [criteria] limits. The
spectrum of hypothetical accidents and transients has been
investigated, and are shown to meet the plant's currently licensed
regulatory criteria. In the area of core design, for example, the
fuel operating limits such as Maximum Average Planar Linear Heat
Generation Rate (MAPLHGR) and Safety Limit Minimum Critical Power
Ratio (SLMCPR) are still met at the uprated power level, and fuel
reload analyses will show plant transients meet the criteria
accepted by the NRC as specified in NEDO-24011, ``GESTAR II''.
Challenges to fuel or ECCS [emergency core cooling system]
performance are evaluated, and shown to still meet the criteria of
10 CFR 50.46 and Appendix K [to 10 CFR 50], (Section 4.3 above, and
Regulatory Guide 1.70 and USAR [Updated Safety Analysis Report]
Section 6.3).
Challenges to the containment have been evaluated, and the
containment and its associated cooling systems will continue to meet
10 CFR 50 Appendix A [General Design Criteria] Criterion 38, Long
Term Cooling, and Criterion 50, Containment.
Radiological release events (accidents) have been evaluated, and
shown to meet the guidelines of 10 CFR 100 (Regulatory Guide 1.70 &
USAR Chapter 15).
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
[[Page 62713]]
As summarized below, this change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Equipment that could be affected by [the] power uprate has been
evaluated. No new operating mode, safety-related equipment lineup,
accident scenario, or equipment failure mode was identified. The
full spectrum of accident considerations defined in Regulatory Guide
1.70 have been evaluated and no new or different kind of accident
has been identified. [The power] Uprate uses already developed
technology, and applies it within the capabilities of already
existing plant equipment in accordance with presently existing
regulatory criteria to include NRC approved codes, standards, and
methods. GE [General Electric] has designed BWRs of higher power
levels than the uprated power of any of the currently operating BWR
fleet and no new power dependent accidents have been identified.
The Technical Specification changes needed to implement [the]
power uprate require some small adjustments, but no change to the
plant's physical configuration. All changes have been evaluated, and
are acceptable.
(3) Will the change involve a significant reduction in a margin
of safety?
The calculated loads on all affected structures, systems and
components will remain within their design allowables for all design
basis event categories. No NRC acceptance criteria will be exceeded.
Only some design and operational margins are affected by [the] power
uprate. The margins of safety originally designed into the plant are
not affected by [the] power uprate. Because the plant configuration
and reactions to transients and hypothetical accidents will not
result in exceeding the presently approved NRC acceptance limits,
[the] power uprate can 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 for the power uprate.
Although not required for the power uprate, the licensee also
requested a change to technical specifications to increase the main
steam SRV setpoint tolerance from +0%, -2% to [plus or minus] 3%.
However, the licensee's no significant hazards consideration for the
power uprate does not expressly address the change to the SRV setpoint
tolerance. Therefore, the NRC staff's review of this change is
presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The main steam SRV's safety function lift setpoints are tested in
accordance with ASME Code requirements and the licensee's inservice
testing program. The setpoint tolerance determines whether the SRV
passes or fails the surveillance requirement and if additional valves
are to be tested. Notwithstanding the results of the safety function
lift setpoint test, if the measured value is outside a tolerance of
[plus or minus] 1%, the valve is reset to within [plus or minus] 1% of
the design lift setpoint. Therefore, the change to the SRV setpoint
tolerance does not affect the performance of any structure, system, or
component in the plant and does not affect the operation of the plant.
Accordingly, the change will not significantly increase the probability
or consequences of an accident previously evaluated.
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The setpoint tolerance change does not alter the function of the
valves' over-pressure protection features, and the release of steam/
water through the SRVs is addressed in previously evaluated accident
analysis. Therefore, the change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
(3) Will the changes involve a significant reduction in a margin of
safety?
The change only affects whether a SRV passes or fails its safety
function surveillance requirement, as well as the total number of
valves to be tested. Regardless the outcome of these tests, all valves
tested will be returned to within [plus or minus] 1% of the design lift
setpoint. The 2% nominal ``as-left'' tolerance span is effectively the
same tolerance span as specified in the current technical
specifications. As a result, there is no significant reduction in a
margin of safety.
Therefore, based on its review, it appears that the three standards
of 10 CFR 50.92(c) are satisfied, and the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 3, 1999.
Description of amendment request: Entergy Operations, Inc.
(licensee) has proposed to revise Final Safety Analysis Report (FSAR)
Section 9.5.4.1, ``Diesel Generator Fuel Oil Storage and Transfer
Systems.'' The revision will change this section of the FSAR to
explicitly list the Waterford Steam Electric Station, Unit 3 (Waterford
3) deviations from the guidance described in American National
Standards Institute (ANSI) N195-1976, ``Fuel Oil Storage System for
Standby Diesel Generator.'' The licensee determined that these proposed
changes require Nuclear Regulatory Commission staff approval prior to
implementation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Waterford 3 FSAR to match the
current design of the Waterford 3 fuel oil storage and transfer
system. The change effectively requests deviations from portions of
ANSI N195-1976. None of these changes significantly increases the
probability of an accident because the Emergency Diesel Generator
(EDG) fuel oil system is not an initiator of any analyzed event.
There are no accidents analyzed in the Final Safety Analysis Report
(FSAR) that are initiated by the systems or components affected by
these changes.
The deviation from ANSI N195-1976, which allows less than the
ANSI Standard recommended volume to be stored in the existing EDG
Fuel Oil Storage Tanks (FOSTs) A and B, will not significantly
increase the consequences of an accident. Waterford 3 contains at
least seven days of fuel oil in each FOST. Although the Waterford 3
FOSTs do not contain a 10% margin, there are numerous diesel fuel
oil vendors nearby from which to obtain fuel oil. Waterford 3 also
has the capability to transport EDG fuel oil from vendors by tanker
truck, train, or barge. This situation ensures that Waterford 3 will
have fuel oil readily available when there is a need for
replenishment. Waterford 3 does not store the additional amount of
fuel oil required for testing. A previous Technical Specification
(TS) Amendment addressed the Waterford 3 FOSTs not containing enough
fuel oil for testing. However, an exception to this requirement was
previously approved in TS Amendment 92.
The request for deviation from the ANSI N195-1976 requirement
for the feed tank suction to be from above the bottom, will not
increase the consequences of any accident. Previous operating
experience at Waterford 3 has shown that since initial startup there
have not been any water or filter blockage problems attributed to
the bottom suction from the feed tank. The fuel oil in each feed
tank is replenished every 31 days during the EDG monthly
Surveillance Requirement (SR). Blockage problems are further
minimized because testing the FOSTs for particulates is performed
with a more conservative filter size than installed on the EDG
engine (0.8
[[Page 62714]]
microns versus 5 microns). Also, TS Surveillances require water and
sediment content to be verified and if water is present, for it to
be removed.
The request for deviation from the ANSI N195-1976 requirement
for the feed tank overflow to discharge to the FOST will not
increase the consequences of any accident. The feed tank is equipped
with design features to ensure fuel oil is not depleted due to over-
filling the feed tank. The feed tank contains a high level switch
that stops the transfer pump upon indication of high level and a
high level alarm that alerts the Control Room of high level in the
tank. A failure of both the feed tank high level switch and high
level alarm occurring simultaneously is very remote. These measures
will not prevent the loss of some fuel oil; however, two failures
would have to occur to prevent the Control Room from being notified.
Even if one EDG FOST were depleted because of the above failures,
the other EDG FOST would be available to ensure seven days of fuel
oil for one EDG.
The request for deviation from the ANSI N195-1976 requirement to
have one pressure indicator located in the discharge of the fuel oil
transfer pump will not increase the consequences of any accident. A
pressure indicator on the discharge of the transfer pump could
indicate performance degradation of the pump; however, the Waterford
3 transfer pumps are designed for automatic operation. If a failure
of the transfer pump occurred, indication would appear in the
Control Room via the alarm for low feed tank level. The alarm for
low feed tank level is adequate to alert the Control Room of a
transfer pump malfunction. If a transfer pump were to malfunction,
the other transfer pump would be available to deliver fuel oil to
operate one EDG for at least seven days. ASME Section XI testing is
performed on the transfer pump once per quarter (temporary pressure
instrumentation is installed on the discharge of the pump to measure
pump differential pressure) to verify that pump performance has not
degraded. In addition, the transfer pumps are functionally tested
every month during routine testing of the EDGs.
The requested deviations from ANSI N195-1976 do not affect the
consequences of an accident because none of the requested deviations
will prevent the EDG from having seven days of fuel oil available
(without multiple failures). Therefore, the EDG fuel oil system will
perform as required to provide sufficient fuel oil to the EDG to
mitigate the consequences of design basis accidents.
Therefore, based on all the above, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
The proposed change revises the Waterford 3 FSAR to match the
current design of the Waterford 3 fuel oil storage and transfer
system. This change is a change to a commitment, and has no [a]ffect
on the current diesel fuel oil storage system or how it is operated,
nor does it [a]ffect any other safety systems or components, or the
way the plant is operated. The change does not affect any accident
analysis assumptions (including a loss of offsite power) or accident
analysis conclusions. Therefore, the proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No
The proposed change revises the Waterford 3 FSAR to match the
current design of the Waterford 3 fuel oil storage system. Although
Waterford 3 deviates from certain ANSI N195-1976 requirements, these
deviations do not result in any changes to the fuel oil storage
system or accident analyses. The deviations do not affect the
ability of any safety systems required to protect the multiple
barriers. No accident mitigatiors are affected by the change because
the amount of available fuel oil has not changed. As a result, the
proposed deviations will not cause a significant decrease in the
margin of safety or prevent Waterford 3 from safely shutting down.
The result of using Probabilistic Safety Assessment techniques
conclude that increasing the fuel oil storage capacity at Waterford
3 to comply with the ANSI requirements has no risk significance. The
specific [a]ffects of the deviations on the margin of safety are
addressed below.
The current TS for stored EDG fuel oil ensures there is
sufficient fuel oil to operate one EDG for seven days assuming the
worst case single active or passive failure. Fuel oil is readily
available due to the number of vendors in the vicinity of Waterford
3. Waterford 3 is also capable of replenishing EDG fuel oil via
tanker truck, train, or barge. Therefore, this change does not
affect the supply of EDG fuel oil being maintained at Waterford 3.
This supply of fuel oil is sufficient to power the ESF systems
required to mitigate design basis accidents. A previous TS Amendment
addressed the Waterford 3 FOSTs not containing enough fuel oil for
testing.
The current feed tank design with the suction from the bottom
instead of on the side as required by ANSI N195-1976 will not
significantly decrease the margin of safety. Waterford 3 has not
experienced particulate or water accumulation in the feed tanks. The
fuel oil in the tank is essentially turned-over every 31 days during
the EDG monthly SR, and TS Surveillances ensure water and sediment
content are verified. Additionally, particulate testing is performed
on the EDG FOSTs using a test filter with a smaller micron size than
is on the engine. This will assure the EDG engine is not subject to
failures due to particulate or water accumulation in the feed tanks.
The request for deviation from the ANSI N195-1976 requirement
for the feed tank overflow to discharge to the FOST will not
significantly decrease the margin of safety. The feed tank is
equipped with two safety measures that would have to fail in order
to allow a loss of EDG fuel oil due to over-filling a feed tank. A
failure of these safety measures (high level switch to stop the
transfer pump and a high level alarm in the feed tank) occurring
simultaneously is very remote.
The request for deviation from ANSI N195-1976 to have one
pressure indicator located at the discharge of the fuel oil transfer
pump will not significantly decrease the margin of safety. A
pressure indicator on the discharge of the transfer pump could
indicate performance degradation of the pump. If a failure of the
transfer pump occurred, indication would appear in the Control Room
via the alarm for low feed tank level. The alarm for low feed tank
is adequate to alert the control room of a transfer pump
malfunction. However, if the transfer pump were to malfunction, the
other transfer pump would be available to deliver fuel oil to
operate one EDG for at least seven days. ASME Section XI testing is
performed on the transfer pump once per quarter (temporary pressure
instrumentation is installed on the discharge of the pump to measure
pump differential pressure) to verify that pump performance has not
degraded. In addition, the transfer pumps are functionally tested
every month during routine testing of the EDGs.
Therefore, based on all the above, the proposed changes 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.
Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: September 27, 1999.
Description of amendment request: The proposed change to the
Technical Specifications (TSs), if approved, will clarify several
administrative requirements, delete redundant requirements, and correct
typographical errors. These revisions affect TS Sections 3.8.3.1,
3.8.3.2, 6.2.2, 6.5.1.2, 6.8.2, 6.9.1.5, and 6.9.1.6.
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 TS changes do not involve a significant increase
in the probability or
[[Page 62715]]
consequences of an accident previously evaluated.
The changes are administrative in nature and do not impact the
operation, physical configuration, or function of plant equipment or
systems. The changes do not impact the initiators or assumptions of
analyzed events, nor do they impact mitigation of accidents or
transient events. Therefore, these changes do not increase the
probability of occurrence or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature and do not
alter plant configuration, require that new equipment be installed,
alter assumptions made about accidents previously evaluated, or
impact the operation or function of plant equipment. Therefore,
these changes do not create the possibility of a new or different
kind of accident than previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature and do not
involve any physical changes to plant structures, systems, or
components (SSCs), or the manner in which these SSCs are operated,
maintained, modified, tested, or inspected. The proposed changes do
not involve a change to any safety limits, limiting safety system
settings, limiting conditions of operation, or design parameters for
any SSC. The proposed changes do not impact any safety analysis
assumptions and do not involve a change in initial conditions,
system response times, or other parameters affecting any accident
analysis. Therefore, these changes do not involve any 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.
Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
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 amendment request: October 1, 1999.
Description of amendment request: The proposed amendments would
revise the minimum fuel oil level for the diesel generator day tanks in
Surveillance Requirement 3.8.1.3 and would change the acceptable fuel
oil level storage band in Required Action Statement B of Limiting
Condition for Operation 3.8.3.
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 diesel generators are designed to supply power to the
emergency systems needed to mitigate the consequences of design
basis accidents such as LOCA/LOSP [loss-of-coolant accident/loss-of-
offsite power]. They (the diesel generators) do not function to
prevent accidents. Reducing the level requirement in the day tanks
and raising the level requirement in the fuel oil storage tanks will
therefore not increase the probability of occurrence of a LOCA/LOSP
event. Furthermore, this proposed change does not affect any other
system or piece of equipment designed to prevent the occurrence of
any other design basis accident or transient. Therefore, reducing
the required level in the day tanks and raising the level in the
fuel oil storage tanks will not increase the probability of
occurrence of any previously evaluated accident or transient.
The consequences of previously evaluated events will not be
significantly increased because, with the 500-gallon day tank
requirement and the increased storage tank supply, ample fuel will
be available to supply the diesel generators for the duration of a
LOCA/LOSP event or a station blackout event. Therefore, the
consequences of an accident previously evaluated are not increased
by this modification.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Lowering TS SR 3.8.1.3 from [greater than or equal to] 900
gallons to [greater than or equal to] 500 gallons and raising TS SR
3.8.3.1 from [greater than or equal to] 33,000 gallons to [greater
than or equal to] 33,320 gallons will have no impact on the normal
or emergency operation of the diesel generator and its support
systems. For example, diesel generator transfer pumps and supply
tank transfer pumps will continue to perform as necessary to insure
an adequate supply in the respective tanks for accident mitigation.
As a result, since no new unanalyzed modes of operation are
introduced, the possibility of a new or different type of accident,
from any previously evaluated is not introduced.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The Bases for TS SR 3.8.1.3 states that the day tank must carry
enough fuel oil to provide for one hour of operation, plus a 10
percent margin. This requirement is based on ANSI N195-1976 (Section
6.1).
The present 900-gallon requirement in the present Technical
Specifications provides for 3.5 hours of continuous operation.
Reducing the volume requirement to 500 gallons will continue to
provide ample margin above the 1-hour requirement. In fact, 500
gallons in the day tank provides for 1.89 hours of continuous
operation.
The Bases for TS SR 3.8.3.1 states that the fuel in the storage
tanks (33,000 gallons) alone is sufficient to account for seven days
of continuous operation. This is true for 33,000 gallons of usable
fuel. However, each storage tank contains approximately 1,438
gallons of unusable fuel. Additionally, part of the current design
bases for the emergency diesel generators is the ability to run four
of the five diesels continuously for seven days at a load of 3250
kW. With 500 gallons in each of the four diesel's day tanks and
33,320 gallons in each of the five storage tanks, the system is
capable of running continuously for 7 days. Ample onsite fuel
capacity remains to operate the diesels continuously for a longer
period than required to replenish the supply from outside sources.
For the above reasons, the margin of safety is not significantly
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.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear
Plant, Unit 1, Appling County, Georgia
Date of amendment request: October 15, 1999.
Description of amendment request: The proposed amendment would
change the Safety Limit Minimum Critical Power Ratios (SLMCPR) in
Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-
specific calculation performed for Unit 1 Operating Cycle 19. The
calculation was done using the new NRC-approved methodology for
determining SLMCPRs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specification changes do not involve a
significant increase in the probability of an accident previously
evaluated.
The derivation of the revised SLMCPRs for Plant Hatch Unit 1
Cycle 19 for incorporation
[[Page 62716]]
into the TS, and their use to determine cycle-specific thermal
limits, have been performed using NRC-approved methods and
procedures. The procedures incorporate cycle-specific parameters and
reduced power distribution uncertainties in the determination of the
lower value for SLMCPRs. These calculations do not change the method
of operating the plant and have no effect on the probability of an
accident initiating event or transient.
The basis of the MCPR Safety Limit is to ensure no mechanistic
fuel damage is calculated to occur if the limit is not violated. The
new SLMCPRs preserve the existing margin to transition boiling and
the probability of fuel damage is not increased. Therefore, the
proposed changes do not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes result only from a revised method of
analysis for the Unit 1 Cycle 19 core reload. These changes do not
involve any new method for operating the facility and do not involve
any facility modifications. No new initiating events or transients
result from these changes. Therefore, the proposed TS changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPRs are calculated using NRC-approved methods and
procedures which are in accordance with the current fuel design and
licensing criteria. The SLMCPRs remain high enough to ensure that
greater than 99.9% of all fuel rods in the core are expected to
avoid transition boiling if the limit is not violated, thereby
preserving the fuel cladding integrity.
Therefore, the proposed TS changes do not involve a reduction in
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.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear
Plant, Unit 2, Appling County, Georgia.
Date of amendment request: October 15, 1999.
Description of amendment request: The proposed amendment would
change the Safety Limit Minimum Critical Power Ratios (SLMCPR) in
Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-
specific calculation performed for Unit 2 Operating Cycle 16. The
calculation was performed using the new NRC-approved methodology for
determining SLMCPRs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specification changes do not involve a
significant increase in the probability of an accident previously
evaluated.
The derivation of the revised SLMCPRs for Plant Hatch Unit 2
Cycle 16 for incorporation into the TS, and their use to determine
cycle-specific thermal limits, have been performed using NRC-
approved methods and procedures. The procedures incorporate cycle-
specific parameters and reduced power distribution uncertainties in
the determination of the lower value for SLMCPRs. These calculations
do not change the method of operating the plant and have no effect
on the probability of an accident initiating event or transient.
The basis of the MCPR Safety Limit is to ensure no mechanistic
fuel damage is calculated to occur if the limit is not violated. The
new SLMCPRs preserve the existing margin to transition boiling and
the probability of fuel damage is not increased. Therefore, the
proposed changes do not involve an increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes result only from a revised method of
analysis for the Unit 2 Cycle 16 core reload. These changes do not
involve any new method for operating the facility and do not involve
any facility modifications. No new initiating events or transients
result from these changes. Therefore, the proposed TS changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPRs are calculated using NRC-approved methods and
procedures which are in accordance with the current fuel design and
licensing criteria. The SLMCPRs remain high enough to ensure that
greater than 99.9% of all fuel rods in the core are expected to
avoid transition boiling if the limit is not violated, thereby
preserving the fuel cladding integrity.
Therefore, the proposed TS changes do not involve a reduction in
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.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: October 18, 1999.
Description of amendment request: The proposed amendment would
revise the activated charcoal testing methodology in accordance with
the guidance provided in NRC Generic Letter 99-02, ``Laboratory Testing
of Nuclear Grade Activated Charcoal.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Will the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
The Standby Gas Treatment (SBGT) system is used to support
mitigation of the consequences of postulated accidents. The SBGT
system is not considered an initiator of any analyzed accident.
There is no change in function or operation of the system. The
proposed change only revises the charcoal laboratory testing
protocol to a more current standard that is more reliable, accurate
and conservative. The change in relative humidity proposed is
likewise in accordance with accepted guidance and reflective of the
Vermont Yankee system configuration, which utilizes heaters to
reduce the incoming humidity. The change in iodide removal
efficiency is also more conservative.
Thus, the probability or consequences of previously analyzed
accidents is not significantly increased.
2. Will the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
[[Page 62717]]
This change does not affect the design or mode of operation of
any plant system, structure or component. No physical alteration of
plant structures, systems or components is involved and no new or
different equipment will be installed. The proposed change only
modifies the laboratory testing protocol and acceptance criteria to
a more currently accepted standard.
Thus, the proposed change does not create the possibility of a
new or different [kind of] accident from those previously evaluated.
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.
The proposed changes in laboratory test protocol do not
adversely affect the operation of any systems, structures or
components. In fact, adopting the newer test standard will provide
greater assurance that the charcoal will perform its intended
function of accident consequence mitigation.
Thus, the proposed change does not significantly reduce 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.
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.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: October 21, 1999.
Description of amendment request: The proposed amendment makes
editorial and administrative changes to the Technical Specifications
(TSs) by correcting two administrative errors and changing the
designation of a TS-referenced figure. These changes do not materially
change the meaning or application of any TS requirement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Will the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed changes are administrative or editorial in nature
and do not involve any physical changes to the plant. The
administrative changes do not materially affect any existing
technical requirement and do not reduce the actions that are
currently taken to ensure operability of plant structures, systems
or components.
The changes correct past administrative errors and change a
reference in the Technical Specifications and do not revise the
methods of plant operation which could increase the probability or
consequences of previously evaluated accidents. No new modes of
operation are introduced by the proposed changes such that a
previously evaluated accident is more likely to occur or more
adverse consequences would result.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
These changes are administrative in nature and do not affect the
operation of any systems or components, nor do they involve any
potential initiating events that would create any new or different
kind of accident. There are no changes to the design assumptions,
conditions, configuration of the facility, or the manner in which
the plant is operated and maintained.
The changes do not affect assumptions contained in plant safety
analyses or the physical design and/or modes of plant operation.
Consequently, no new failure mode is introduced due to the
administrative changes.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated for Vermont Yankee.
3. Will the proposed changes involve a significant reduction in
a margin of safety?
There are no changes being made to the Technical Specification
safety limits or safety system settings. The operating limits and
functional capabilities of systems, structures, and components are
unchanged as a result of these administrative changes. These
proposed changes do not affect any equipment involved in potential
initiating events or safety limits. There is no change to the basis
for any Technical Specification that is related to the establishment
of, or the maintenance of, a nuclear safety margin.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
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.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: October 5, 1999.
Description of amendment request: The proposed amendments would
make changes to the Technical Specifications (TSs) that are necessary
to eliminate inconsistencies in the TSs pertaining to decay heat
removal requirements (TSs 15.3.1.A.3, 15.3.3.A, and 15.3.3.C). An
additional change to the requirements in TS 15.3.1.A.4 for pressurizer
safety valve operability is also proposed to provide appropriate
coordination with low temperature overpressure protection requirements.
Bases revisions are provided consistent with the proposed amendments
and to administratively correct references related to accumulator
operability in the Bases for TS 15.3.3.
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 Point Beach Nuclear Plant in accordance with
the proposed amendments will not create a significant increase in
the probability or consequences of an accident previously evaluated.
Technical Specifications 15.3.1.A.3, 15.3.3.A.3 and 15.3.3.C are
all interrelated in that they each provide direction for required
decay heat removal capability, either directly or indirectly by
providing requirements for both support and supported systems. TS
15.3.1.A.3 provides requirements for the operation of the reactor
coolant system loops, steam generators, reactor coolant pumps and
residual heat removal loops as necessary to support decay heat
removal from a shutdown unit. TS 15.3.3.A provides requirements for
operation of the high head safety injection and low head residual
heat removal system. Specifically, TS 15.3.3.A.3 provides
requirements for inoperability of the residual heat removal system
which accounts for the dual purpose of injection and decay heat
removal. TS 15.3.3.C.2 provides requirements for operation of the
Component Cooling Water System, a primary support system for both
Residual Heat Removal System and Reactor Coolant Pump operation. The
proposed Specifications require redundancy of decay heat removal and
require placing the plant in a safe condition, maximizing the
availability of decay heat removal methods when redundancy is lost.
Appropriate allowances and actions are required to ensure uniform
mixing of boron for reactivity control with the unit shutdown and
provide for appropriate allowances to facilitate surveillance
testing, and refueling operations. The time limits placed on all
actions are consistent with safe operations, industry and NRC
guidance. Therefore the probability of a
[[Page 62718]]
loss of shutdown cooling or loss of subcooling; or a loss of
shutdown reactivity control is minimized.
Amendments are also proposed to provide for coordination of
Pressurizer Safety Valve and Pressurizer Power Operated Relief Valve
operability requirements to ensure redundant overpressure protection
is provided for all operating conditions. Proposed actions for
inoperability of Pressurizer Safety Valves minimizes the time in
that condition. Operation of the valves is not changed. Thus, the
probability of a loss of coolant due to inadvertent opening of the
valves is not increased. In addition, overpressure protection is
maintained under all conditions such that the probability of an
overpressure due to an analyzed event is not increased.
The proposed changes do not affect potential leakage paths for
radiation to the environment, or of key safety barriers, and ensure
appropriate system and function redundancy is maintained. Therefore
the consequences of an accident previously evaluated will not
increase.
Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendments does not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendments do not alter the operation or method of
function of the Residual Heat Removal System, Component Cooling
Water System, Pressurizer Safety Valves, or Power Operated Relief
Valves. The amendments provide for consistency of decay heat removal
and pressure relief requirements within the Specifications providing
assurance these functions can be maintained during all required
plant conditions. Operations are not altered in any way that could
introduce a new accident initiator not previously considered in the
PBNP Safety Analyses. Therefore, operation of the Point Beach
Nuclear Plant in accordance with the proposed amendments cannot
create the possibility of a new or different kind of accident than
any previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant reduction
in a margin of safety.
The proposed amendments ensure redundancy of the decay heat
removal and overpressure protection over the complete range of
operating conditions. Limitations are provided to ensure timely
action to restore the functions to an operable condition consistent
with their importance to safety. Appropriate allowances and actions
are required to ensure uniform mixing of boron for reactivity
control with the unit shutdown and provide for appropriate
allowances to facilitate surveillance testing, and refueling
operations consistent with overall safety. The functions or method
of function of the systems or components affected are not being
altered. Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendments cannot result in a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three 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.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 21, 1999.
Description of amendment request: The request proposes to revise
Technical Specification (TS) 3.4.10, Pressurizer Safety Valves (PSV),
of the improved Technical Specifications issued March 31, 1999. The
proposed revision is to reduce the safety valve set pressure in
Limiting Condition for Operation (LCO) 3.4.10, and increase the
setpoint tolerance in Surveillance Requirement (SR) 3.4.10.1. The PSV
setpoint and setpoint tolerance is proposed to be changed from 2485
psig plus or minus 1% to 2460 psig plus or minus 2% in the LCO. The
tolerance of plus or minus 1% in the SR is for resetting the setpoint
after testing, if this is needed. The licensee also submitted the Bases
pages for TS 3.4.10, which show modifications to reflect the changes to
the TSs.
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.
Any evaluations performed on an overpressure transient
conservatively assume the upper limit of the pressurizer safety
valve (PSV) tolerance as the pressure to which the reactor coolant
system (RCS) is subjected. The proposed change to the lower
tolerance limit of the pressure set point means that an overpressure
transient may be terminated at a pressure that is lower than assumed
in the analysis. It has also been determined that the design
transients are not adversely affected because the limiting
transients are not sensitive to the pressure tolerance decrease.
Therefore, the primary system pressure boundary is not challenged by
the PSV lower tolerance limit change. The change in the upper limit
of the PSV tolerance does not challenge the upper limit of the
overpressure protection. The maximum opening set pressure is not
changed, and therefore, does not impact analyses performed for
overpressure transients. Although the lower PSV set point would
result in a lower qualified valve flow rate, the slightly lower
valve flow rate would be more than compensated for by the reduced
valve opening pressure. The change to the PSV set point and set
point tolerance does not change the conclusions of the existing
thermal hydraulic analysis for the pressurizer safety and relief
system. The design function of the valves is not being changed.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated in the USAR [Wolf Creek Updated Safety Analysis
Report].
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change would allow the PSV minimum actuation
pressure to be as low as 2411 psig. The pressurizer power-operated
relief valve (PORV) actuation set point is 2335 psig. Therefore, the
margin between the PORV and PSV actuation set points could be as low
as 76 psi, which is a reduction of 49 psi from the current 125 psi
margin. Even with the 30 psi pressure control uncertainty, the
actuation set point margin of 76 psi is considered adequate and the
PORVs are expected to continue to actuate before the PSVs during
Condition 1 transients. As such, the proposed change will not have
any adverse effect on the control systems. Except for the reduced
lower set point, the design and operation of the PSVs are not being
changed. The maximum opening pressure is not being changed. The only
effect of this change would be that the PSVs could open at a lower
pressure, but still above the PORV actuation set point. Therefore,
the possibility of a new or different kind of accident from any
accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The PSVs provide, in conjunction with the reactor protection
system, overpressure protection for the RCS. The PSVs are designed
to prevent the system pressure from exceeding the system safety
limit, 2735 psig, which is 110% of the design pressure. The change
in the upper limit of the PSV tolerance from plus or minus 1% to
plus or minus 2% with a reduction in the nominal set point from 2485
psig to 2460 psig does not challenge the upper limit of the
overpressure protection. The maximum opening pressure set point is
not changed, and therefore, does not impact analyses performed for
overpressure transients. The change to PSV set point and set point
tolerance does not change the conclusions of the existing thermal
hydraulic analysis for the pressurizer safety and relief system. For
all non-LOCA [non-loss of coolant accident] events the analyses
support the change in PSV set point and set point tolerance from
2485 psig plus or minus 1% to 2460 psig plus or minus 2%. The change
in the PSV set
[[Page 62719]]
point and set point tolerance also has no effect on the Reactor
Protection or Engineered Safety Features Systems trip set points.
Thus, the proposed change does not involve a significant reduction
in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Indiana Michigan Power Company, Docket No. 50-315 and 50-316, Donald C.
Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: September 23, 1999, as supplemented
October 11, 1999.
Brief description of amendment request: The proposed amendments
involve movement of loads in excess of the design-basis seismic
capability of the auxiliary building load handling equipment and
structures. The proposed amendment requests approval to move the steam
generator sections through the auxiliary building and to disengage
crane travel interlocks, and also requests relief from performance of
Technical Specification Surveillance Requirement 4.9.7.1.
Date of publication of individual notice in Federal Register:
October 26, 1999 (64 FR 57665).
Expiration date of individual notice: November 26, 1999.
Indiana Michigan Power Company, Docket No. 50-315 and 50-316, Donald C.
Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: October 1, 1999.
Brief description of amendment request: The proposed amendments
involve the resolution of an unreviewed safety question related to
certain small-break loss-of-coolant accident scenarios for which there
may not be sufficient containment recirculation sump water inventory to
support continued operation of the emergency core cooling system and
containment spray system pumps during and following switchover to cold
leg recirculation. Resolution of this issue consists of a combination
of physical plant modifications, new analyses of containment
recirculation sump inventory, and resultant changes to the accident
analyses to ensure sufficient water inventory in the containment
recirculation sump. In addition, the licensee proposes to change the
Technical Specifications dealing with the refueling water storage tank
inventory and temperature, the required amount of ice in each ice
basket in the containment, and the delay to start the containment air
recirculation/ hydrogen skimmer fans.
Date of publication of individual notice in Federal Register:
October 29, 1999 (64 FR 58458).
Expiration date of individual notice: November 29, 1999.
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 electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
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: April 21, 1999, as supplemented
October 15, 1999.
Brief description of amendment: The amendment allows for a one-time
extension of the reactor protection system and engineered safety
features actuation system instruments.
Date of issuance: October 29, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 205.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes October 14, 1999 (64 FR 55777). The October
15, 1999, letter provided clarifying information that did not change
the initial proposed no significant hazards consideration. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided for an opportunity to request a hearing by October 28, 1999,
but indicated that if the Commission makes a final NSHC determination,
any such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of NSHC are contained in
a Safety Evaluation dated October 29, 1999.
[[Page 62720]]
Attorney for licensee: Mr. Brent L. Brandenburg, Assistant General
Counsel, Consolidated Edison Company of New York, Inc., 4 Irving
Place--1822, New York, NY 10003.
NRC Section Chief: Sheri Peterson.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: August 4, 1999.
Brief description of amendments: The amendments revise the TS
(Appendix A of the Catawba operating licenses) to: (1) modify Section
3.3.2 regarding the Nuclear Service Water System, and (2) Section 5.3.1
regarding operating personnel qualifications.
Date of issuance: November 2, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: Unit 1-181; Unit 2-173.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48861).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 1999.
No significant hazards consideration comments received: No
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: October 22, 1997.
Brief description of amendment: This amendment approves a proposed
modification that changes the Perry facility as described in the
Updated Safety Analysis Report. The change incorporates temperature
control valves and associated bypass lines around the Emergency Closed
Cooling system heat exchangers. These features are designed to ensure
operability of the Control Complex Chilled Water System under post-
accident load conditions, without the need for compensatory measures.
Date of issuance: October 29, 1999.
Effective date: October 29, 1999.
Amendment No.: 107.
Facility Operating License No. NPF-58: This amendment authorizes
the revision of the Updated Safety Analysis Report.
Date of initial notice in Federal Register: November 5, 1997 (62 FR
59922).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 29, 1999.
No significant hazards consideration comments received: No
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: May 18, 1999, as supplemented
by letter dated September 22, 1999.
Brief description of amendments: The amendments revised
Surveillance Requirements (SR) 3.8.1.3 and 3.8.1.13 to reduce the
loading requirements for the emergency diesel generators (EDGs).
Revised SR 3.8.1.3 requires the EDGs be loaded and operated for
[greater than or equal to] 60 minutes at a load [greater than or equal
to] 6500 kW and [less than or equal to] 7000 kW at least every 31 days.
Revised SR 3.8.1.13 requires the EDGs to be loaded [greater than or
equal to] 6900kW and [less than or equal to] 7700 kW and operated as
close as practicable to 3390 kVA for 2 hours. For the remaining hours
of the test, the EDGs would be loaded [greater than or equal to] 6500
kW and [less than or equal to] 7000 kW and operated as close as
practicable to 3390 kVA.
Date of issuance: October 25, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-109; Unit 2-87.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43780) The supplemental letter dated September 22, 1999, provided
clarifying information that did not change the scope of the May 18,
1999, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 25, 1999.
No significant hazards consideration comments received: No
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: April 28, 1999.
Brief Description of amendments: These amendments revise TS Section
3.4.A.4 for Units 1 and 2. The changes relax the minimum volume
requirement for the refueling water Chemical Addition Tank (CAT) from
4200 gallons to 3930 gallons. A minor administrative change is also
being made to TS Table 4.1-2B to correct an earlier printing error and
to delete a reference which no longer applies.
Date of issuance: November 1, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 222 and 222.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48869).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 1, 1999.
No significant hazards consideration comments received: No
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a
[[Page 62721]]
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
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room).
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By December 17, 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 electronically from the ADAMS Public Library
component on the NRC Web site, http://www.nrc.gov (the Electronic
Reading Room). If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, 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
[[Page 62722]]
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).
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: March 26, 1999, as supplemented
October 15, 1999.
Brief description of amendment: The amendment allows for a one-time
extension of system functional tests. The test intervals are extended
for 37 months to coincide with the next refueling outage scheduled to
commence on June 3, 2000.
Date of issuance: October 29, 1999.
Effective date: As of the date of issuance to be implemented upon
receipt.
Amendment No.: 204.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Press release issued requesting comments as to proposed no
significant hazards consideration: Yes, October 22 and 24, 1999,
Peekskill Evening Star.
The October 15, 1999, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration. The notice provided an opportunity to submit comments on
the Commission's proposed NSHC determination. No comments have been
received. The notice also provided for an opportunity to request a
hearing by October 28, 1999, but indicated that if the Commission makes
a final NSHC determination, any such hearing would take place after
issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of NSHC are contained in
a Safety Evaluation dated October 29, 1999.
Attorney for licensee: Mr. Brent L. Brandenburg, Assistant General
Counsel, Consolidated Edison Company of New York, Inc., 4 Irving
Place--1822, New York, NY 10003 NRC Section Chief: Sheri Peterson.
Dated at Rockville, Maryland, this 9th day of November 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management Office of Nuclear
Reactor Regulation.
[FR Doc. 99-29846 Filed 11-16-99; 8:45 am]
BILLING CODE 7590-01-P
