[Federal Register Volume 68, Number 130 (Tuesday, July 8, 2003)]
[Notices]
[Pages 40707-40725]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-17028]
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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, June 13, 2003, through June 26, 2003. The
last biweekly notice was published on June 24, 2003 (68 FR 37574).
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
[[Page 40708]]
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 Administrative 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 Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By August 7, 2003, 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 PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent 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
[[Page 40709]]
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 PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: June 2, 2003.
Description of amendment request: The licensee proposed to revise
Sections 3.7.B.1 and 3.7.C.2 of the OCNGS Technical Specifications
(TSs). Section 3.7.B.1 currently specifies that the reactor may remain
in operation ``for a period not to exceed 7 days in any 30 day period
if a startup transformer is out of service.'' Section 3.7.C.2,
referring to the standby diesel generators (DGs), currently specifies
that the reactor may remain in operation ``for a period not to exceed 7
days in any 30 day period if a diesel generator is out of service.''
The proposed revision is to delete the phrase ``in any 30 day period''
from these two sections. The licensee regards this phrase as an
unnecessary restriction, and states that it has no basis in the
existing TSs, design basis, or licensing basis of OCNGS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has reviewed the licensee's analysis against the three
standards of 10 CFR 50.92(c) and performed its own. The NRC staff's
analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed changes, if approved by the NRC staff, will be
made in a manner such that conservatism is maintained through continued
compliance with applicable NRC regulations (specifically, the
Maintenance Rule in 10 CFR 50.65) and guidance. No hardware design
change is involved with the proposed amendment, thus there can be no
adverse effect on the functional performance of the startup
transformers or DGs. Consequently, the subject components will continue
to perform their design functions with no decrease in their
capabilities to mitigate the consequences of postulated accidents.
Unavailability of these components was not factored into the scenarios
of previously analyzed accidents, nor were the subject components
assumed to be initiators of previously analyzed accidents.
Consequently, the proposed revision to the subject sections will lead
to no increase in the consequences of accidents previously evaluated,
and will lead to no increase of the probability of accidents previously
evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods OCNGS is operated. As a
result, all structures, systems, and components will continue to
perform as previously analyzed by the licensee, and previously
evaluated and accepted by the NRC staff. Therefore, the proposed
amendment will not create the possibility of a new or different kind of
accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, the proposed amendment
will not affect in any way the performance characteristics and intended
functions of the subject components. Therefore, the proposed amendment
does not involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, 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 E. Matthews, Esquire, Morgan, Lewis, &
Bockius, LLP, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Richard J. Laufer.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: May 28, 2003.
Description of amendments request: The amendments would modify
several surveillance requirements (SRs) in Technical Specifications
(TSs) 3.8.1 and 3.8.4 on alternating current and direct current
sources, respectively, for plant operation. The revised SRs would have
notes deleted or modified to allow the SRs to be performed, or
partially performed, in reactor modes that are currently not allowed by
the TSs. The current SRs are not allowed to be performed in Modes 1 and
2. Several of the current SRs also cannot be performed in Modes 3 and
4. The footnote to SR 3.8.4.8 would also be deleted. There would also
be renumbering in several of the SR notes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The emergency diesel generators (DGs) and their associated
emergency loads are accident mitigating features, rather than
accident initiating equipment. Each DG is dedicated to a specific
vital bus and these buses and DGs are independent of each other.
There is no common mode failure provided by the testing changes
proposed in this license amendment request (LAR) that would cause
multiple bus failures. Therefore, there will be no significant
impact on any accident probabilities by the approval of the
requested amendment.
The design of plant equipment is not being modified by these
proposed changes. The changes include an increase in the online time
the DG will be paralleled to the grid in Mode 1, 2, 3, [and] 4. The
overall time that the DG is paralleled in all modes (outages/non-
outage) should remain unchanged. As such, the ability of the DGs to
respond to a design basis accident (DBA) can be adversely impacted
by [the] proposed changes. However, the impacts are not considered
significant based on the DG under test maintaining its ability to
respond to an auto-start signal were one to be received during
testing, along with the ability of the remaining DG to mitigate a
DBA or provide a safe shutdown, and data that shows that the DG
itself will not perturb the electrical system significantly.
Furthermore, the
[[Page 40710]]
proposed amendments for surveillance requirements (SR) 3.8.1.10 and
SR 3.8.1.14 share the same electrical configuration alignment to the
current monthly 1-hour loaded surveillance.
For SR 3.8.1.13, the DG would still be able to respond to an
auto-start signal were one to be received during testing. The
unavailability of the DG during the conduct of this SR 3.8.1.13 is
minimal (approx[imately] 30 minutes) and is considered insignificant
from a risk perspective.
In addition, operating experience and evaluation of the
probability of a DG being rendered inoperable concurrent with or due
to a significant grid disturbance, support the conclusion that the
proposed changes in this LAR do not involve any significant increase
in the likelihood of a safety-related bus blackout.
SR changes that are consistent with Industry/Technical
Specification Task Force (TSTF) Standard Technical Specification
(STS) change TSTF-283, Revision 3 and NUREG-1432, Revision 2 have
been approved by the NRC, and the on-line tests allowed by the TSTF
and the NUREG are only to be performed for the purpose of
establishing operability [of the DG being tested]. Performance of
these SRs during previously restricted modes will require an
assessment to assure plant safety is maintained or enhanced.
The deletion of the footnote associated with SR 3.8.4.8 is an
editorial change. This footnote was associated with coming out of
the ninth refueling outage for Unit 1, which has since passed.
Therefore, the proposed change[s do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different [kind of] accident from any accident previously
evaluated.
The proposed change[s] would create no new accidents since no
changes are being made to the plant that would introduce any new
accident causal mechanisms. Equipment will be operated in the same
configuration currently allowed by other DG SRs that allow testing
in plant Modes 1, 2, 3, and 4. This license amendment request does
not impact any plant systems that are accident initiators or
adversely impact any accident mitigating systems.
Therefore, the proposed change[s do] not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not involve a significant reduction in
the margin of safety. The margin of safety is related to the ability
of the fission product barriers to perform their design [safety]
functions during and following an accident situation. These barriers
include the fuel cladding, the reactor coolant system, and the
containment system. The proposed changes to the testing requirements
for the plant DGs do not affect the operability requirements for the
DGs, as verification of such operability will continue to be
performed as required (except during different allowed modes [of
operation]). Continued verification of operability supports the
capability of the DGs to perform their required function of
providing emergency power to plant equipment that supports or
constitutes the fission product barriers. Only one DG is to be
tested at a time and the remaining DG will be available to safely
[shut down] the plant or respond to a DBA, if required.
Consequently, the performance of these fission product barriers will
not be impacted by implementation of [the] proposed amendment.
In addition, the proposed changes involve no changes to [safety]
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
Therefore, the proposed change[s do] not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Stephen Dembek.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: May 28, 2003.
Description of amendments request: The proposed amendment would
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,
cooldown curves (Technical Specification Figure 3.4.3-2) to change the
range of temperatures for which a cooldown rate of 100 [deg]F/hr is
acceptable.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
In accordance with 10 CFR Part 50, Appendix G, the Calvert
Cliffs pressure/temperature (P-T) limits for material fracture
toughness requirements of the reactor coolant pressure boundary
materials were developed using the methods of linear elastic
fracture mechanics and the guidance found in the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section
III, Appendix G. The proposed cooldown rates for the Technical
Specification P-T limits were made possible by ASME Code Case N-640
which permits use of KIC for reference stress intensity
factor. [Temperatures that enable the low temperature overpressure
protection system are not affected].
The proposed change only changes the temperature at which the
cooldown transitions from 100[deg]F/hr to 40[deg]F/hr. It does not
change the basic cooldown rates or methods of cooling down the
Reactor Coolant System. This cooldown transition does not affect the
probability of an accident previously evaluated because the cooldown
rates have not changed. Additionally, since the cooldown rates are
not changed above 300[deg]F, the safety analyses and dose
consequences in the Updated Final Safety Analysis Report are not
affected.
Therefore[,] the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The implementation of the proposed revision has no significant
effect on either the configuration of the plant, or the manner in
which it is operated.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety is defined by compliance with 10 CFR Part
50, Appendix G, requirements for adequate margin to prevent brittle
failure of the reactor coolant pressure boundary materials. As
discussed above, use of KIC with continuous cooldown
results in a conservative cooldown rate that will maintain plant
safety. With the proposed change, the underlying intent of the 10
CFR Part 50, Appendix G, is maintained.
Therefore, this 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
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: May 28, 2003.
Description of amendment request: The proposed amendment would
modify the Technical Specifications requirements for spent fuel storage
pool
[[Page 40711]]
boron concentration and fuel storage. The proposed amendment would
eliminate the need to credit Boraflex neutron absorbing material for
reactivity control in the H. B. Robinson Steam Electric Plant, Unit No.
2, spent fuel storage pool. The new analyses submitted by the licensee
take credit for a combination of soluble boron and controlled fuel
loading patterns within the spent fuel storage pool in order to
maintain acceptable margins of subcriticality.
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:
An evaluation of the proposed change has been performed in
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards
considerations using the standards in 10 CFR 50.92(c). A discussion
of these standards as they relate to this amendment request follows:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed changes do not modify the facility. They apply
additional administrative controls for maintaining the required
boron concentration in the spent fuel storage pool. They also revise
the acceptance criteria for the spent fuel storage pool criticality
analyses. There will be a procedural change requiring increased
frequency of spent fuel storage pool sampling for boron analysis.
The sampling is performed in accordance with approved procedures and
does not impact the probability or consequences of spent fuel
storage pool accidents, which are a fuel handling accident and a
loss of spent fuel storage pool cooling. The changes will allow for
the further degradation of the Boraflex within the high density
racks. The existence or degradation of the Boraflex has no
relationship to the probability or consequences of a fuel handling
accident or a loss of spent fuel storage pool cooling.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
The proposed changes are related to the possibility of a
criticality accident in the spent fuel storage pool. Detailed
analyses have been performed to ensure a criticality accident in the
spent fuel storage pool is not a credible event. The events that
could lead to a criticality accident are not new. These events
include a fuel mis-positioning event, a fuel drop event, and a boron
dilution event. The proposed changes do not impact the probability
of any of these events. The detailed criticality analyses performed
demonstrate that criticality would not occur following any of these
events. For the more likely events, such as a fuel mis-positioning
event, keff remains less than or equal to 0.95. For the
unlikely event that the spent fuel storage pool boron concentration
was reduced to zero, keff remains less than 1.0. Since a
criticality accident remains ``not credible,'' the proposed changes
do not create the possibility of a new or different kind of accident
from any previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
The proposed changes continue to provide the controls necessary
to ensure a criticality event could not occur in the spent fuel
storage pool. The acceptance criteria are consistent with the
acceptance criteria specified in 10 CFR 50.68, which provide an
acceptable margin of safety in regard to the potential for a
criticality event. Therefore, the changes do not result in a
significant reduction in the margin of safety.
Based on the above discussion, [Carolina Power & Light Company]
has determined that the requested change does not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen G. Howe.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: March 4, 2003, as supplemented May 13,
2003.
Description of amendment request: The proposed amendment would
revise selected sections of the Technical Specifications (TSs) based
upon a re-analysis of fuel handling accidents (FHAs). The revised
analysis is based upon selective implementation of the alternative
source term (AST) methodology of Regulatory Guide (RG) 1.183, and in
accordance with Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.67. Specifically, the amendment would revise: TS 3.7.8,
``Plant Systems, Control Room Envelope Pressurization System;'' TS
3.9.4, ``Refueling Operations, Containment Building Penetrations;'' TS
3.9.9, ``Refueling Operations, Containment Purge and Exhaust Isolation
System,'' and TS 3.9.12, ``Refueling Operations, Fuel Building Exhaust
Filter System.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The staff's review is
presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes do not involve physical modifications to the
plant equipment and do not change the operational methods or procedures
used for the physical movement of fuel in containment or in the fuel
building. As such, the proposed changes have no effect on the
probability of occurrence of any accident previously evaluated.
The proposed changes are based upon the re-analysis of an FHA in
the containment and an FHA in the fuel building area. The consequences
of the re-analyzed events are expressed in terms of total effective
dose equivalent (TEDE), and are not directly comparable to either the
thyroid or whole body doses reported in the existing analyses. However,
even taking this comparison into consideration, any dose increase is
considered not to be significant as the revised analyses results meet
the applicable TEDE acceptance criteria for AST implementation.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The containment closure components (e.g., equipment access hatch,
personnel access hatch doors, and various containment penetrations) and
filtration systems are not accident initiators. The proposed changes do
not involve the addition of new systems or components nor do they
involve the modification of existing plant systems. The proposed
changes do not change the operational modes or procedure used for the
physical movements of fuel in containment or in the fuel building. The
proposed changes do not affect the way in which an FHA is postulated to
occur. Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The margin of safety for the dose consequence analysis is
considered to be that provided by meeting the applicable regulatory
limits. The dose consequences of the existing FHA are within regulatory
limits for whole body and thyroid doses as established in 10 CFR 100.
The revised FHA using the AST method demonstrates that the dose
consequences are within the regulatory limits for TEDE established in
10 CFR 50.67 and RG 1.183. There is no direct
[[Page 40712]]
correlation between the old margins of safety established by meeting 10
CFR Part 100 and those established by the proposed change. The staff
concludes, however, that meeting 10 CFR 50.67 and RG 1.183 limits would
result in doses that would be within the 10 CFR Part 100 limits.
Therefore, it is concluded that a reducation in margin of safety, if
any, would not be significant.
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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: James W. Clifford.
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: February 25, 2003, as supplemented June
9, 2003.
Description of amendment request: The proposed amendments require a
Steam Generator (SG) Program that defines a performance based approach
to maintaining SG tube integrity. The SG Program includes performance
criteria that define the basis for tube integrity and provides
reasonable assurance that SG tubing will remain capable of fulfilling
its safety function of maintaining reactor coolant pressure boundary
(RCPB) integrity. The proposed amendments add a new Technical
Specification (TS) for SG Tube Integrity (3.4.18) and revise the TSs
for Reactor Coolant System (RCS) Operational Leakage (3.4.13), SG Tube
Surveillance Program (5.5.9), and SG Tube Inspection Report (5.6.8).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the facility in accordance with the proposed
amendments:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes require a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown, and all anticipated transients included in the
design specification) and design basis accidents.
The SG performance criteria are based on tube structural
integrity, accident induced leakage, and operational leakage.
The structural integrity performance criterion is a new
requirement. It is included in the proposed SG Program
administrative TS 5.5.9.
The accident induced leakage criterion is a new requirement. It
is included in the proposed SG Program administrative TS 5.5.9.
The operational leakage criterion is equivalent to the existing
requirement. Its limit is part of the proposed RCS Operational
Leakage TS 3.4.13.
A SG tube rupture event is one of the design basis accidents
analyzed as part of Catawba's licensing basis. In the analysis of a
SG tube rupture event, a bounding primary to secondary leakage rate
equal to the operational leakage rate limit in the licensing basis
plus the leakage rate associated with a double-ended rupture of a
single tube is assumed. For other design basis accidents, the tubes
are assumed to retain their structural integrity (i.e., they are
assumed not to rupture). These analyses assume that primary to
secondary leakage through each SG is 150 gallons per day.
The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed as part of these TS amendments
identify the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining RCPB integrity throughout
each operating cycle and in the unlikely event of a design basis
accident. The performance criteria are only a part of the SG Program
required by the proposed changes to TS 5.5.9. The program, defined
by NEI [Nuclear Energy Institute] 97-06, ``Steam Generator Program
Guidelines,'' includes a framework that incorporates a balance of
prevention, inspection, evaluation, repair, and leakage monitoring.
Probability of an Accident
The TS proposed by these license amendments define the actions
required upon failure to maintain SG tube integrity and the
surveillances necessary to verify that tube integrity is maintained.
The proposed administrative TS contain performance criteria, repair
criteria, repair methods, maximum SG inspection intervals, and
reporting requirements. The set of TS proposed is a significant
improvement over the existing SG TS.
In addition, the SG Program required by these amendments
includes provisions important in satisfying the TS requirements. The
topics addressed by the SG Program include:
[sbull] SG performance criteria, including an operational
leakage limit,
[bull] SG repair criteria and repair methods,
[bull] SG inspection intervals, and
[sbull] Performance based SG inspections that include pre-
inspection degradation assessments, condition monitoring
assessments, operational assessments, and non-destructive
examination technique requirements.
These SG Program provisions establish requirements that are an
improvement as compared to the requirements in the existing TS. As
an example, the SG Program requires an operational assessment that
defines the maximum SG inspection interval that provides reasonable
assurance that the performance criteria will continue to be met at
the next inspection. The actual inspection interval is always chosen
to be less than the interval determined by the operational
assessment. The existing TS have no similar requirement. As a
result, the function and integrity of the tubes are maintained with
greater assurance and the probability of a SG tube rupture is
decreased.
Consequences of an Accident
The consequences of design basis accidents are, in part,
functions of the dose equivalent I131 in the primary
coolant and the primary to secondary leakage rates resulting from an
accident. Therefore, limits are included in the plant TS for
operational leakage and for dose equivalent I131 in
primary coolant to ensure the plant is operated within its analyzed
condition.
The analysis of the associated design basis accidents assumes
that the initial primary to secondary leak rate is 150 gallons per
day in each SG (except for the ruptured SG in a SG tube rupture),
and that the reactor coolant activity levels of dose equivalent
I131 are at the TS values before the accident. The TS
limits, license conditions, and other controls on I131
are unchanged by these amendment requests. These other controls
include License Amendments 159 and 151 for Catawba Units 1 and 2,
respectively, and the Catawba license amendment request submittal
dated May 9, 2002, which is presently being reviewed by the NRC.
In addition, the proposed amendments include a new performance
criterion for accident induced leakage that requires that the
primary to secondary leakage resulting from an accident other than a
SG tube rupture not exceed the value assumed in the dose analyses
(150 gallons per day through each SG).
Since the proposed operational leakage limit is equivalent to
the existing value, and since the proposed amendments include a new
performance criterion for accident induced leakage, the proposed
amendments will not increase the consequences of an accident.
From the above discussion, it is concluded that the proposed
amendments do not affect the design of the SGs, their method of
operation, or primary coolant chemistry controls. The proposed
approach updates the existing TS and enhances the requirements for
SG inspections. The proposed TS changes do not adversely impact any
other previously evaluated design basis accident and represent an
improvement over the existing TS. Therefore, the proposed changes do
not affect the consequences of a SG tube rupture accident and the
probability of such an accident is reduced. In addition, the
proposed changes do not affect the consequences of other accidents.
[[Page 40713]]
2. Would not create the possibility of a new or different kind
of accident from any other accident previously evaluated.
The proposed performance based requirements are an improvement
over the requirements imposed by the existing TS. Implementation of
the proposed SG Program will not introduce any adverse changes to
the plant design basis or postulated accidents resulting from
potential tube degradation. The result of the implementation of the
SG Program will be an enhancement of SG tube performance. Primary to
secondary leakage that may be experienced during all plant
conditions will be monitored to ensure it remains within current
accident analysis assumptions.
The proposed amendments do not affect the design of the SGs,
their method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed changes do not impact any other
plant system or component. The changes enhance SG inspection
requirements. Therefore, the proposed changes do not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The SG tubes in pressurized water reactors are an integral part
of the RCPB and, as such, are relied upon to maintain the primary
system's pressure and inventory. As part of the RCPB, the SG tubes
are unique in that they are also relied upon as a heat transfer
surface between the primary and secondary systems such that residual
heat can be removed from the primary system. In addition, the SG
tubes also isolate the radioactive fission products in the primary
coolant from the secondary system. In summary, the safety function
of a SG is maintained by ensuring the integrity of its tubes.
SG tube integrity is a function of the design, environment, and
physical condition of the tube. The proposed license amendments do
not affect tube design or operating environment. The proposed
changes are expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the existing TS.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed revisions to
the TS.
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 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: May 21, 2003.
Description of amendment request: Change the technical
specifications by extending the functional test frequency of the
reactor protection system (RPS) intermediate range monitor (IRM)
functions from weekly to 31 days, and to add more restrictive
requirements for the RPS IRM--High Flux function.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Section 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration. The NRC
staff has reviewed the licensee's analysis against the standards of 10
CFR 50.92(c). The staff's review is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously analyzed?
The proposed changes do not physically impact the plant, nor do
they impact any design or functional requirements of the associated
systems. The change does not degrade the performance of, or increase
the challenges to, any safety systems assumed to function in the safety
analysis. The changes do not impact the way in which surveillances are
performed or introduce any accident initiators. The availability of
equipment and systems required to prevent or mitigate the radiological
consequences of an accident are not significantly affected because of
other, more frequent testing that is performed, the availability of
redundant systems and equipment, or the high reliability of the
equipment. More stringent requirements that ensure operability of
equipment do not affect the initiation of any event, nor do they
negatively impact the mitigation of any event.
Therefore, the proposed changes do 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 changes do not introduce any failure mechanisms of a
different type than previously evaluated, since no physical changes to
the plant are being made. No new failure modes are introduced as no new
or different equipment is being installed, and no installed equipment
is being operated or surveillance tested in a different manner.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in the margin of
safety?
Although the proposed changes would result in changes to the
interval between certain surveillance tests, the impact, if any, on
system availability is minimal, based upon other more frequent testing
that is performed, the existence of redundant systems and equipment, or
overall system reliability. The changes do not significantly impact the
condition or performance of structures, systems, and components relied
upon for accident mitigation. The imposition of more stringent
requirements has no negative impact on margins of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
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.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: May 30, 2003.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TS) and other elements
of the licensing bases to maintain a Post Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the TS
[[Page 40714]]
for nuclear power reactors currently licensed to operate. Lessons
learned and improvements implemented over the last 20 years have shown
that the information obtained from PASS can be readily obtained through
other means or is of little use in the assessment and mitigation of
accident conditions.
The changes are based on NRC-approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System
(PASS).'' The NRC staff issued a notice of opportunity for comment in
the Federal Register on December 27, 2001 (66 FR 66949), on possible
amendments concerning TSTF-413, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on March 20, 2002 (67 FR 13027). The licensee affirmed the
applicability of the following NSHC determination in its application
dated May 30, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a
PASS provides little actual benefit to post accident mitigation.
Past experience has indicated that there exists in-plant
instrumentation and methodologies available in lieu of a PASS for
collecting and assimilating information needed to assess core damage
following an accident. Furthermore, the implementation of Severe
Accident Management Guidance (SAMG) emphasizes accident management
strategies based on in-plant instruments. These strategies provide
guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management
strategies and guidelines, it is determined that the PASS provides
little benefit to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from TS (and
other elements of the licensing bases) does not involve a
significant increase in the consequences of any accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. Jonathan Rogoff, General Counsel,
Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: June 3, 2003.
Description of amendment request: The proposed amendment would
revise the Palisades Plant Operating License and Technical
Specifications to increase the licensed rated power level by 1.4
percent from 2530 megawatts thermal (MWt) to 2565.4 MWt. This power
level increase is considered a measurement uncertainty recapture power
uprate.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed increase in power level is achieved by the taking
credit for the accuracy of the existing feedwater flow measurement
instrumentation, including the Crossflow ultrasonic flow measurement
(UFM) system, which results in a more accurate feedwater flow used
in the heat balance calculation. The increased flow accuracy
utilizing the Crossflow UFM system improves the uncertainty in the
core power level from the existing 2 percent margin to <= 0.5925%.
The probability of an accident previously evaluated is not increased
by the proposed change because the flow measurement instrumentation
is not an initiator of design-basis accidents evaluated in the
updated final safety analysis report [FSAR].
The plant design and licensing basis has been evaluated for
operation at the proposed increased value of 2565.4 Megawatts
thermal (MWt). All systems and components continue to acceptably
perform their structural and operational functions.
There are no changes as a result of the proposed measurement
uncertainty recapture power uprate to the design or operation of the
plant that could affect system, component, or accident mitigative
functions. All systems and components will function as designed and
the applicable performance requirements have been evaluated and
found to be acceptable. The proposed variable high power trip
allowable value will ensure that the maximum actual steady state
power at
[[Page 40715]]
which a trip would be actuated is within safety analysis limits.
Therefore, there is no significant increase in the probability
of an accident previously evaluated.
The reduction in power measurement uncertainty is bounded by the
safety analyses since they were performed or evaluated at 2580.6
MWt. Radiological consequences of [FSAR] Chapter 14 accidents were
assessed previously and continue to be bounding. The FSAR Chapter 14
analyses continue to demonstrate compliance with the relevant
accident analysis acceptance criteria. Therefore, there is no
significant increase in the consequences of any accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed change. All
systems, structures and components previously required for the
mitigation of an event remain capable of fulfilling their intended
design function at the proposed uprated power level. The proposed
change has no adverse effects on any safety-related systems or
component and does not challenge the performance or integrity of any
safety-related system. The proposed variable high power trip
allowable value will ensure that the maximum actual steady state
power at which a trip would be actuated is within safety analysis
limits. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The maximum steady-state reactor power of 2580.6 MWt assumed in
the accident analysis, including uncertainties, remains the same as
previously analyzed. Therefore, the change in rated thermal power to
2565.4 MWt does not involve a significant reduction in the margin of
safety.
The current accident analyses and system and component analyses
had been previously performed at core powers that exceed the
proposed measurement uncertainty recapture uprated core power.
Evaluations have been performed for analyses that were done at
nominal core power and have been found acceptable for the proposed
measurement uncertainty recapture power uprate. Analyses of the
primary fission product barriers at uprated core powers have
concluded that all relevant design basis criteria remain satisfied
in regard to integrity and compliance with the regulatory acceptance
criteria. As appropriate, all evaluations have been either reviewed
and approved by the Nuclear Regulatory Commission or are in
compliance with applicable regulatory review guidance and standards.
The proposed variable high power trip allowable value will ensure
that the maximum actual steady state power at which a trip would be
actuated is within safety analysis limits. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: L. Raghavan.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: May 29, 2003.
Description of amendment requests: The License Amendment Request
(LAR) revises TS 3.8.1, ``AC Sources--Operating'' to allow surveillance
testing of the onsite standby emergency diesel generators (DG) during
modes in which it is currently prohibited. Specifically, the licensee
proposes removing the mode restrictions for the following surveillance
requirements (SRs): SR 3.8.1.10 (full load rejection test), SR 3.8.1.13
(protective-trip bypass test), and SR 3.8.1.14 (endurance and margin
test). This LAR also incorporates changes included in the NRC-approved
Industry/Technical Specification Task Force (TSTF) Standard Technical
Specification (STS) change TSTF-283, Revision 3. These changes modify
the Notes in SRs 3.8.1.8 (transfer of AC sources test), 3.8.1.9 (post
accident load rejection test), 3.8.1.11 (simulated loss of offsite
power test), 3.8.1.12 (auto-start on safety injection (SI) signal
test), 3.8.1.16 (restoration of loads to offsite power test), 3.8.1.17
(verification of test mode override test), 3.8.1.18 (engineered safety
feature and auto-transfer load sequencing test), 3.8.1.19 (loss of
offsite power plus SI signal response test), 3.8.4.7 (battery service
test), and 3.8.4.8 (battery discharge test) to allow performance of the
surveillances in order to reestablish operability following corrective
maintenance, corrective modification, deficient or incomplete
surveillance testing, and other unanticipated operability concerns
during plant operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The emergency diesel generators (DGs) and their associated
emergency loads are accident-mitigating features. As such, testing
of the DGs themselves is not associated with any potential accident
initiating mechanism. Each DG is dedicated to a specific vital bus
and these buses and DGs are independent of each other. There is no
common mode failure provided by the testing changes proposed in this
license amendment request (LAR) that would cause multiple bus
failures. Therefore, there will be no significant impact on any
accident probabilities by the approval of the requested amendment.
The design of plant equipment is not being modified by these
proposed changes.
The changes include an increase in the online time the DG will
be paralleled to the grid in Mode 1 or 2. However, the overall time
that the DG is paralleled in all modes (outage/non-outage) should
remain unchanged. As such, the ability of the DGs to respond to a
design basis accident can be adversely impacted by these proposed
changes. However, the impacts are not considered significant based
on the ability of the remaining two DGs to mitigate a design bases
accident (DBA) or provide a safe shutdown, and data that shows that
the DG itself will not perturb the electrical system. Furthermore,
the proposed amendments for surveillance requirement (SR) 3.8.1.10
and SR 3.8.1.14 share the same electrical configuration alignment to
the current monthly 1-hour loaded surveillance.
For SR 3.8.1.13, the DG would still be able to respond to an
auto-start signal were one to be received during testing. The
unavailability of the DG during the conduct of this SR 3.8.1.13 is
minimal (approximately 5 minutes) and is insignificant from a risk
perspective.
In addition, operating experience and evaluation of the
probability of a DG being rendered inoperable concurrent with or due
to a significant grid disturbance support the conclusion that the
proposed changes in this LAR do not involve any significant increase
in the likelihood of a safety-related bus blackout.
SR changes that are consistent with Industry/Technical
Specification Task Force (TSTF) Standard Technical Specification
(STS) change TSTF-283, Revision 3 have been approved by the NRC and
the online tests allowed by the TSTF are only to be performed for
the purpose of establishing operability. Performance of these SRs
during normally restricted modes will require an assessment to
assure plant safety is maintained or enhanced.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different accident from any accident previously evaluated.
The proposed change would create no new accidents since no
changes are being made to the plant that would introduce any new
accident causal mechanisms. Equipment will be operated in the same
configuration currently allowed by other DG SRs that allow
[[Page 40716]]
testing in plant Modes 1 and 2 and 3. This license amendment request
does not impact any plant systems that are accident initiators or
adversely impact any accident mitigating systems.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change does not involve a significant reduction in
the margin of safety. The margin of safety is related to 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. The proposed changes to the testing requirements for the
plant DGs do not affect the operability requirements for the DGs, as
verification of such operability will continue to be performed as
required (except during different allowed modes). Continued
verification of operability supports the capability of the DGs to
perform their required function of providing emergency power to
plant equipment that supports or constitutes the fission product
barriers. Consequently, the performance of these fission product
barriers will not be impacted by implementation of this proposed
amendment.
In addition, the proposed changes involve no changes to
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: June 5, 2003.
Brief description of amendments: The proposed change involves the
extension from 1 hour to 24 hours of the completion time (CT) for
Condition B of Technical Specification (TS) 3.5.1, which defines
requirements for accumulators. Accumulators are part of the emergency
core cooling system and consist of tanks partially filled with borated
water and pressurized with nitrogen gas. The contents of the tank are
discharged to the reactor coolant system (RCS) if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the
accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore
an accumulator to operable status when it has been declared inoperable
for a reason other than the boron concentration of the water in the
accumulator not being within the required range. This change was
proposed by the Westinghouse Owners Group participants in the Technical
Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is
supported by NRC-approved topical report WCAP-15049-A, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,'' submitted
on May 18, 1999. The NRC staff issued a notice of opportunity for
comment in the Federal Register on July 15, 2002 (67 FR 46542), on
possible amendments concerning TSTF-370, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on March 12, 2003 (68 FR 11880). The licensee affirmed the
applicability of the following NSHC determination in its application
dated June 5, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR)
[[Page 40717]]
correlation limit, the design DNBR limits, or the safety analysis
DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable
accumulator for up to 24 hours, instead of 1 hour, before the plant
would be required to begin shutting down. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk. Since
the frequency of a design basis large LOCA (a large LOCA with loss
of offsite power) would be significantly lower than the large LOCA
frequency of the WCAP-15049-A evaluation, the impact of increasing
the accumulator CT from 1 hour to 24 hours on plant risk due to a
design basis large LOCA would be significantly less than the plant
risk increase presented in the WCAP-15049-A evaluation.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: June 11, 2003.
Description of amendment requests: The license amendment request
proposes to revise Technical Specification (TS) 3.1.7, ``Rod Position
Indication,'' TS 3.2.1, ``Heat Flux Hot Channel Factor,'' TS 3.2.4,
``Quadrant Power Tilt Ratio,'' and TS 3.3.1, ``Reactor Trip System
Instrumentation,'' to allow use of a power distribution monitoring
system as described in WCAP-12472-P-A, ``BEACON Core Monitoring and
Operations Support System,'' for power distribution measurements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The power distribution monitoring system (PDMS) performs
continuous core power distribution monitoring. This system utilizes
the NRC-approved Westinghouse proprietary computer code, the Best
Estimate Analyzer for Core Operations--Nuclear (BEACON), to provide
data reduction for incore flux maps, core parameter analysis, load
follow operation simulation, and core prediction. It in no way
provides any protection or control system function. Fission product
barriers are not impacted by these proposed changes. The proposed
changes occurring with PDMS will not result in any additional
challenges to plant equipment that could increase the probability of
any previously evaluated accident. The changes associated with the
PDMS do not affect plant systems such that their function in the
control of radiological consequences is adversely affected. These
proposed changes will therefore not affect the mitigation of the
radiological consequences of any accident described in the Final
Safety Analysis Report Update (FSARU).
Continuous on-line monitoring through the use of PDMS provides
significantly more information about the power distributions present
in the core than is currently available. This results in more time
(i.e., earlier determination of an adverse condition developing) for
operator action prior to having an adverse condition develop that
could lead to an accident condition or to unfavorable initial
conditions for an accident.
Each accident analysis addressed in the Diablo Canyon Power
Plant FSARU is examined with respect to changes in cycle-dependent
parameters, which are obtained from application of the NRC-approved
reload design methodologies, to ensure that the transient evaluation
of reload cores are bounded by previously accepted analyses. This
examination, which is performed in accordance with the requirements
set forth in 10 CFR 50.59, ``Changes, tests and experiments,''
ensures that future reloads will not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The implementation of the PDMS has no influence or impact on
plant operations or safety, nor does it contribute in any way to the
probability or consequences of an accident. No safety-related
equipment, safety function, or plant operation will be altered as a
result of this proposed change. The possibility for a new or
different type of accident from any accident previously evaluated is
not created since the changes associated with implementation of the
PDMS do not result in a change to the design basis of any plant
component or system. The evaluation of the effects of using the PDMS
to monitor core power distribution parameters shows that all design
standards and applicable safety criteria limits are met.
The proposed changes do not result in any event previously
deemed incredible being made credible. Implementation of the PDMS
will not result in more adverse conditions and will not result in
any increase in the challenges to safety systems. The cycle specific
variables required by the PDMS are calculated using NRC-approved
methods. The Technical Specifications will continue to require
operation within the required core operating limits and appropriate
actions will be taken when or if limits are exceeded.
The proposed change, therefore, does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is not affected by the implementation of
the PDMS. The margin of safety provided by current TS remains
unchanged. The proposed changes continue to require operation within
the core limits that are based on NRC-approved reload design
methodologies. Appropriate measures exist to control the values of
these cycle-specific limits. The proposed changes continue to ensure
that appropriate actions will be taken if limits are violated. These
actions remain unchanged.
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 requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PSEG Nuclear, LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: June 17, 2003.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) Surveillance Requirement
4.6.2.1.b.2.b. This change would remove the requirement to verify that
the reactor thermal power output is less than, or equal to, 1% of rated
thermal power when the suppression chamber average water temperature is
above 95 [deg]F. Additionally, the amendment would correct two
typographical errors on TS index page ``x.''
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Section
[[Page 40718]]
50.91(a), the licensee has provided its analysis of the issue of no
significant hazards consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not affect the allowable suppression
chamber average water temperatures provided in the TS. The changes
do not affect previously evaluated events described in the UFSAR
[Updated Final Safety Analysis Report] including all DBAs [Design
Basis Accidents] and other operational transients.
The surveillance is extraneous because Action b of LCO [Limiting
Condition for Operation] 3.6.2 directs the plant operators to
commence a plant shutdown if the suppression chamber temperature
cannot be restored. These changes do not affect plant systems,
structures or components (SSCs).
Therefore, the proposed changes do not involve a significant
increase in the probability or radiological consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not affect the design function or
operation of a plant SSC. No physical or procedural changes are
associated with this LCR [License Change Request]. As a result, no
new credible failure mechanisms, malfunctions, or accident
initiators are related to this change. Additionally, no new modes of
plant operation are created.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes include the deletion of a surveillance
requirement. This change is prompted by an LCO action statement,
which prevents the plant from performing the surveillance. As a
result, this change does not impact safety margins specified in the
Hope Creek licensing basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: May 21, 2003.
Description of amendment request: The proposed amendment would
change the source term for the Dose Calculation Methodology to the
Alternate Source Term (AST). This change would result in design
modifications to the Control Room Emergency Air Treatment System
(CREATS), eliminate the requirement for the Containment Post Accident
Charcoal Filters, and revise both the reactor coolant dose equivalent
I-131 specific activity limit and the containment spray NaOH
concentration limit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The function of the CREATS is to provide a safe environment for
the operators in the event of an accident, and thereby allow them to
perform their accident mitigation responsibilities. The physical
changes to the CREATS were designed to enhance the ability of the
system to perform that function. The new system is an improvement in
reliability, redundancy and leak tightness over the existing system.
The change in design has no impact on accident initiation
frequencies. Therefore the physical changes to the plant do not
increase the probability or consequences of a previously evaluated
accident.
The proposed Technical Specification changes involving the
CREATS reflect the new system configuration and current industry
guidance. The specifications ensure system functionality and
protection of the operators under postulated accident conditions.
The new dose analysis indicates that the radiation dose to the
operators and the public is acceptable without crediting the post
accident charcoal filters removed from Technical Specification 3.6.6
and 5.5.10, and also bounds the change to the Reactor Coolant System
activity limits in Technical Specification 3.4.16. The change to the
dose conversion factor definition in Technical Specification
[S]ection 1.1 is consistent with the new analysis.
The reference to ICRP-30 [International Commission on
Radiological Protection Publication No. 30] in the Dose Equivalent
I-131 definition is consistent with the new analysis and Standard
Tech Specs, NUREG1431, [``Standard Technical Specifications
Westinghouse Plants.'']
All calculated doses are within the regulatory limits prescribed
in 10CFR50.67. In addition, with the exception of one calculated
Exclusion Area Boundary (EAB) dose, all dose numbers are within the
guidelines of Reg Guide 1.183, [``Alternative Radiological Source
Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors,''] and Standard Review Plan (SRP) 15.0.1. This above-
mentioned dose is in one particular direction from the source. The
associated accident is the Locked Rotor Accident, which was not
previously evaluated for dose at Ginna. The 100% fuel failure
assumption used in this accident is widely considered to be overly
conservative. Additionally, extra margin is built into the
calculation because RG&E [Rochester Gas & Electric Corporation]
assumed 500 gallons per day (GPD) of Steam Generator (SG) tube
leakage per SG. Since the primary release pathway for this accident
is SG tube leakage, and Reg Guide 1.183 (reference 3) allows an
assumed tube leakage equal to the Tech Spec allowable leakage
([sim]150 GPD/SG at Ginna), RG&E assumed a release rate of [sim]3.3
times greater than required. The calculated dose (2.7 Rem) is well
below the regulatory limit of 25 Rem and only slightly greater than
the published guideline of 2.5 Rem. Given the localized nature,
associated probability/risk, and conservatism in this analysis, the
calculated dose is considered acceptable.
Iodine removal was not credited in the existing analysis of
doses for Equipment Qualification. Therefore, even though the
Containment Post Accident Charcoal Filters will be removed from Tech
Specs as a result of this amendment, it is not necessary to re-
analyze these doses.
The Toxic Gas in-leakage analysis is bounded by the assumed in-
leakage in the dose analysis. The amendment also does not hinder or
change the ability to mitigate smoke infiltration as described in
NEI [Nuclear Energy Institute] 99-03, Control Room Habitability
Guidance.
This change has no impact on accident initiators, will not
affect the ability of the operators to perform their designated
functions, and removal of the requirement for CNMT [Containment]
Post Accident Charcoal Filters is shown to be acceptable. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
For the proposed changes, a different kind of accident would
involve a situation where the operators would become incapacitated
or otherwise be prevented from fulfilling their function. The new
system differs in that the cooling in the emergency mode is from
direct expansion of R-22 refrigerant. A rupture of the coils could
introduce the refrigerant into the Control Room environment.
However, the charge of refrigerant R-22 in cooling system will be
limited such that a rupture in the cooling coils would not exceed
nationally accepted toxicity standards.
The radiation and/or toxic gas exposures are shown to be
acceptable, and the ability
[[Page 40719]]
of the plant to mitigate smoke infiltration has not changed. The new
system will improve the environmental conditions in most situations
and actually enhance the ability of the operators to perform their
functions.
Given the above, an event that would result in preventing the
operators from fulfilling their safety functions is not introduced
by 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. Does the proposed change involve a significant reduction in
margin of safety?
Response: No.
The new analysis was performed without crediting the existing
Containment Post Accident Charcoal Filters and indicated that the
Control Room and off-site doses remain within the required limits.
Removal of the Post Accident Charcoal Filters from Technical
Specification will not impact the operators' ability to function or
significantly increase dose to the public.
The new Technical Specification surveillance limits for NaOH
tank level and concentration establish criteria acceptable to meet
the assumptions in the dose analysis.
The changes to the VFTP [Ventilation Filter Testing Program]
program in Technical Specification reflect the removal of the
Containment Post Accident Charcoal Filters consistent with the
analysis, and the surveillance limits consistent with the new CREATS
design.
The use of AST represents a change to a standardized and
accepted dose calculation method.
The function of the CREATS system is to protect the operators
and allow them to perform the necessary accident mitigation tasks.
The proposed changes to the CREATS enhance this ability through
improved redundancy and system operation. The analysis demonstrates
that the Control Room will remain within prescribed limits during
the design basis accidents. The operators will be able to perform
their function and the public will be protected.
Therefore, the proposed change does not involve a significant
reduction in a margin to 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 22, 2003 (TSC 03-02).
Description of amendment request: The proposed amendment would
revise the limiting condition for operation for Technical Specification
(TS) Section 3.7.5, ``Ultimate Heat Sink.'' This revision would modify
the required minimum ultimate heat sink (UHS) water elevation in TS
3.7.5.a from 670 feet to 674 feet. The maximum emergency raw cooling
water (ERCW) temperature requirement in TS 3.7.5.b will be increased
from 83 degrees Fahrenheit ([deg]F) to 87 [deg]F. Limiting condition
for operation requirements that are now obsolete because of the
proposed changes are being deleted, as well as expired footnote
provisions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change to increase the UHS maximum temperature
and the minimum water level does not alter the function, design, or
operating practices for plant systems or components. The UHS is
utilized to remove heat loads from plant systems during normal and
accident conditions. This function is not expected or postulated to
result in the generation of any accident and continues to adequately
satisfy the associated safety functions with the proposed changes.
Therefore, the probability of an accident presently evaluated in the
safety analyses will not be increased because the UHS function does
not have the potential to be the source of an accident and no plant
equipment is altered as a result of this change. The heat loads that
the UHS is designed to accommodate have been evaluated for
functionality with the higher temperature and elevation
requirements. The result of these evaluations is that there are
existing margins associated with the systems that utilize the UHS
for normal and accident conditions. These margins are sufficient to
accommodate the postulated normal and accident heat loads with the
proposed changes to the UHS. Since the safety functions of the UHS
are maintained, the systems that ensure acceptable offsite dose
consequences will continue to operate as designed. Therefore, the
proposed changes to TS 3.7.5 will not significantly increase the
consequences of an accident previously evaluated based on safety
functions continuing to meet their accident mitigation requirements
and limiting dose consequences to acceptable levels.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The UHS function is not an initiator of any accident and
only serves as a heat sink for normal and upset plant conditions. By
allowing the proposed change in the UHS temperature and elevation
requirements, only the parameters for UHS operation are changed
while the safety functions of the UHS and systems that transfer the
heat sink capability continue to be maintained. The UHS function
provides accident mitigation capabilities and does not reflect the
potential for accident generation. Therefore, the possibility for
creating a new or different kind of accident is not created because
the UHS is only utilized for heat removal functions that are not a
potential source for accident generation.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed change has been evaluated for systems that are
needed to support accident mitigation functions as well as normal
operational evolutions. Operational margins were found to exist in
the systems that utilize the UHS capabilities such that these
proposed changes will not result in the loss of any safety function
necessary for normal or accident conditions. The ERCW system has
excess flow margins that will accommodate the increased flows
necessary for the proposed temperature increase. While operating
margins have been reduced by the proposed changes, safety margins
have been maintained as assumed in the accident analyses for
postulated events. Additionally, the proposed changes do not require
the modification of component setpoints or operating provisions that
are necessary to maintain margins of safety established by the SQN
design. Therefore, a significant reduction in the margin to safety
is not created by this proposed change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: June 5, 2003 (TSC 03-07).
Description of amendment request: The proposed amendment would
revise Action b of Technical Specification (TS) 3.6.1.9, ``Containment
Ventilation System'' to allow an alternative to returning the
inoperable containment purge supply or exhaust valve to operable
conditions for continued operation. The alternative ensures isolation
of the affected flow path such that potential release paths to the
environment are sufficiently restricted to meet regulatory limits. This
change
[[Page 40720]]
will minimize the need to initiate a unit shutdown or delay start-up
when acceptable means are available to ensure the required safety
function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change does not alter any plant system or
operating practice. This change will allow the isolation of the
affected flow path such that the safety function is completed when
the associated automatic isolation valve is inoperable because of
leakage. The containment purge supply and exhaust valves are not
considered to be the source of an accident as their function is to
isolate containment from the outside environs in the event of an
accident. Accident generation probability is not affected by
providing alternative isolation methods that continue to satisfy the
required safety function. Therefore, the proposed change does not
involve an increase in the probability of an accident previously
evaluated.
The proposed addition for the isolation of the affect flow path
in place of a required shutdown of the unit, provides an equivalent
safety function without the risk associated with a unit shutdown.
Using a feature that has minimal potential for inadvertent loss of
function and a more frequent surveillance to ensure that the
isolation function is maintained, is as good or better than the
automatic system that is required by the TSs. This is because the
proposed action utilizes a passive feature in place of an active
system and ensures offsite dose consequences within required limits.
Additionally, the overall plant safety is enhanced by not requiring
a unit shutdown when acceptable measures can be taken to preserve
the safety function of the containment purge supply and exhaust
valves. Therefore, the proposed change does not involve an increase
in the consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change does not involve a change to plant
systems, components, or operating practices that could result in a
change in accident generation potential. In addition, the purge and
exhaust valves are utilized for the isolation of flow paths to the
environs and are not a feature that could generate a postulated
accident. Use of the proposed action for inoperable purge and
exhaust valves will not impact the potential for accidents.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed changes do not alter plant systems or their
setpoints that are used to maintain the margin of safety.
Additionally, the proposed change provides a method to ensure the
safety function of the containment ventilation and isolation systems
are retained for accident mitigation purposes. The proposed change
will improve the margin of safety by not requiring a unit shutdown
when acceptable methods for maintaining plant safety functions can
be achieved. Therefore, the proposed change does not involve a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
(WBN) Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: May 30, 2003.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to replace the single boron
concentration requirement with a table that defines the minimum and
maximum amount of boron that is required for accident mitigation based
on the number of tritium producing burnable absorber rods (TPBARs) in
the core.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the required boron concentration
for the cold leg accumulators (CLAs) and RWST [Refueling Water
Storage Tank]. The proposed values have been verified to maintain
the required accident mitigation safety function for the CLAs and
RWST. The CLAs and RWST safety function is to mitigate accidents
that require the injection of borated water to cool the core and to
control reactivity. These functions are not potential sources for
accident generation and the modification of the boron concentration
that supports event mitigation will not increase the potential for
an accident. Therefore, the possibility of an accident is not
increased by the proposed changes. The boron levels for this change
are based on the number of TPBARs in the core. As the number of rods
is increased the need for additional shutdown boron also increases.
This effect has been evaluated with a similar methodology utilized
for previously NRC approved amendments associated with tritium
production. This methodology ensures that the impact of TPBARs is
adequately compensated for by the required boron concentrations and
has been incorporated into the proposed revision. Since the boron
levels will continue to maintain the safety function of the CLAs and
RWST in the same manner as currently approved, the consequences of
an accident are not increased by the proposed changes.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change only modifies boron concentrations for
accident mitigation functions of the CLAs and RWST. These functions
do not have a potential to generate accidents as they only serve to
perform mitigation functions associated with an accident. The
proposed requirements will maintain the mitigation function in an
identical manner as currently approved. There are no plant equipment
or operational changes associated with the proposed revision other
than the adjustment of the boron level in the CLAs and RWST.
Therefore, since the CLA and RWST functions are not altered and the
plant will continue to operate without change, the possibility of a
new or different kind of an accident is not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This change proposes boron concentration requirements that
support the accident mitigation functions of the CLAs and RWST
equivalent to the currently approved limits. The proposed change
does not alter any plant equipment or components and does not alter
any setpoints utilized for the actuation of accident mitigation
system or control functions. The proposed boron values have been
verified to provide an adequate level of reactivity control for
accident mitigation. Therefore, the proposed change 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
[[Page 40721]]
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: June 5, 2003.
Brief description of amendments: The proposed change involves the
extension from 1 hour to 24 hours of the completion time (CT) for
Condition B of Technical Specification (TS) 3.5.1, which defines
requirements for accumulators. Accumulators are part of the emergency
core cooling system and consist of tanks partially filled with borated
water and pressurized with nitrogen gas. The contents of the tank are
discharged to the reactor coolant system if, as during a loss of
coolant accident, the coolant pressure decreases to below the
accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore
an accumulator to operable status when it has been declared inoperable
for a reason other than the boron concentration of the water in the
accumulator not being within the required range. This change was
proposed by the Westinghouse Owners Group participants in the Technical
Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is
supported by NRC-approved topical report WCAP-15049-A, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,'' submitted
May 18, 1999. The NRC staff issued a notice of opportunity for comment
in the Federal Register on July 15, 2002 (67 FR 46542), on possible
amendments concerning TSTF-370, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on March 12, 2003 (68 FR 11880). The licensee affirmed the
applicability of the following NSHC determination in its application
dated June 5, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable
accumulator for up to 24 hours, instead of 1 hour, before the plant
would be required to begin shutting down. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk. Since
the frequency of a design basis large LOCA (a large LOCA with loss
of offsite power) would be significantly lower than the large LOCA
frequency of the WCAP-15049-A evaluation, the impact of increasing
the accumulator CT from 1 hour to 24 hours on plant risk due to a
design basis large LOCA would be significantly less than the plant
risk increase presented in the WCAP-15049-A evaluation.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: June 6, 2003.
Brief description of amendments: The proposed license amendments
would change Technical Specification (TS) Section 3.8.4, ``DC Sources--
Operating,'' TS Section 3.8.5, ``DC Sources--Shutdown,'' and TS Section
3.8.6, ``Battery Cell Parameters,'' and
[[Page 40722]]
add a new TS Section 5.5.19, ``Battery Monitoring and Maintenance
Program'', to establish an administrative controls program for the
maintenance and monitoring of the station safety-related batteries. The
purpose of the proposed changes is to provide increased operational
flexibility and allow more efficient application of plant resources to
safety significant activities.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change affects TS sections 3.8.4 ``DC Sources--
Operating,'' TS 3.8.5 ``DC Sources--Shutdown,'' TS 3.8.6 ``Battery
Cell Parameters,'' and TS Administrative Controls section 5.5.
The proposed change restructures the TS for the DC electrical
power subsystem and adds new Conditions and Required Actions with
increased Completion Times to address battery charger inoperability.
Neither the DC electrical power subsystem nor associated battery
chargers are initiators of any accident sequence analyzed in the
updated Final Safety Analysis Report (FSAR). Operation in accordance
with the proposed TS ensures that the DC electrical power subsystem
is capable of performing its function as described in the FSAR,
therefore the mitigative functions supported by the DC electrical
power subsystem will continue to provide the protection assumed by
the analysis.
The relocation of preventive maintenance surveillance, and
certain operating limits and actions to a newly-created, licensee-
controlled TS [5.5.19], ``Battery Monitoring and Maintenance
Program,'' will not challenge the ability of the DC electrical power
subsystem to perform its design function. The maintenance and
monitoring required by current TS, which are based on industry
standards, will continue to be performed. In addition, the DC
electrical power subsystem is within the scope of 10 CFR 50.65,
``Requirements for monitoring the effectiveness of maintenance at
nuclear power plants,'' which will ensure the control of maintenance
activities associated with the DC electrical power subsystem.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any physical alteration of
the units. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints at which protective or mitigative actions are
initiated that are affected by the proposed changes. The operability
of the DC electrical power subsystem in accordance with the proposed
TS is consistent with the initial assumptions of the accident
analyses and is based upon meeting the design basis of the plant.
These proposed changes will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No alteration in the procedures, which ensure
the unit remains within analyzed limits, is proposed, and no change
is being made to procedures relied upon to respond to an off-normal
event. As such, no new failure modes are being introduced. The
proposed changes do not alter assumptions made in the safety
analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not adversely affect operation of plant
equipment and will not result in a change to the setpoints at which
protective actions are initiated. Sufficient DC capacity to support
operation of mitigation equipment is ensured. The changes associated
with the new Battery Maintenance and Monitoring Program will ensure
that the station batteries are maintained in a highly reliable
manner. The equipment fed by the DC electrical system will continue
to provide adequate power to safety related loads in accordance with
analysis assumptions.
Therefore, the proposed changes do 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: June 9, 2003.
Description of amendment request: The proposed amendment would make
administrative changes to Section 6 of the Surry Power Station
Technical Specifications (TS) for Units 1 and 2 to adopt the format for
topical report references that are described in Industry/Technical
Specifications Task Force Traveller, TSTF-363, Rev 0, ``Revised Topical
Report References in Improved Technical Specification (ITS) 5.6.5,
COLR.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed change is administrative in nature and as such does
not impact the condition or performance of any plant structure,
system or component. The proposed administrative change does not
affect the initiators of any previously analyzed event or the
assumed mitigation of accident or transient events. As a result, the
proposed change to the Surry Technical Specifications does not
involve any increase in the probability or the consequences of any
accident or malfunction of equipment important to safety previously
evaluated since neither accident probabilities nor consequences are
being affected by this proposed administrative change.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed change is administrative in nature, and therefore
does not involve any changes in station operation or physical
modifications to the plant. In addition, no changes are being made
in the methods used to respond to plant transients that have been
previously analyzed. No changes are being made to plant parameters
within which the plant is normally operated or in the setpoints,
which initiate protective or mitigative actions, and no new failure
modes are being introduced. Therefore, the proposed administrative
change to the Surry Technical Specifications does not create the
possibility of a new or different kind of accident or malfunction of
equipment important to safety from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed change is administrative in nature and does not
impact station operation or any plant structure, system or component
that is relied upon for accident mitigation. Furthermore, the margin
of safety assumed in the plant safety analysis is not affected in
any way by the proposed administrative change. Therefore, the
proposed change to the Surry Technical Specifications does 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: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion
[[Page 40723]]
Resources Services, Inc., Millstone Power Station, Building 475, 5th
Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
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, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: November 16, 2001, as
supplemented by letters dated October 4, 2002, and March 28, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.6.2.2, ``Suppression Pool Water Level,'' and TS
3.6.2.4, ``Suppression Pool Makeup System,'' to permit draining the
reactor cavity pool portion of the upper containment pool in MODE 3,
``Hot Shutdown,'' with the reactor vessel pressure less than 235 psig.
Date of issuance: June 12, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 156.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15621). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register Notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 12, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendment request: June 26, 2002, as supplemented November
22, 2002.
Brief description of amendment: The amendments change the Technical
Specifications for the pressure-temperature limits curves in Technical
Specification 3.4.9, ``RCS Pressure and Temperature (P/T) Limits.''
Date of issuance: June 18, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance
Amendment No.: 228 and 256.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50949). The November 22, 2002, supplement contained clarifying
information only and did not change the initial no significant hazards
consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 18, 2003.
No significant hazards consideration comments received: No.
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 application of amendments: October 24, 2002, as
supplemented by letters dated November 21, 2002, and February 19, 2003.
Brief description of amendments: The amendments revise TS 3.5.3,
Low Pressure Injection, Condition A, to change the Completion Time from
72 hours to 7 days. This revision will allow longer corrective
maintenance to be completed at power, without requiring a plant
shutdown. It will also reduce shutdowns due to a Limiting Condition for
Operation requirement.
Date of Issuance: June 18, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 332, 332, and 333.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78517).
The supplement dated November 21, 2002, did not change the scope of
the October 24, 2002, application; however it did change the licensee's
proposed No Significant Hazards Consideration Determination (NSHCD).
The supplement dated February 19, 2003, provided clarifying information
that did not change the scope of the October 24, 2002, application nor
the initial proposed NSHCD.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 18, 2003.
No significant hazards consideration comments received: No.
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: March 19, 2003.
Brief description of amendment: The amendment deletes Technical
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminates the requirements to have and maintain the post accident
sampling system at River Bend Station, Unit 1. The amendment also
addresses related changes to TS 5.5.2, ``Primary Coolant Sources
Outside Containment.''
[[Page 40724]]
Date of issuance: June 23, 2003.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 134.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 29, 2003 (68 FR
22746).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 23, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania
Date of application for amendment: July 24, 2002, as supplemented
February 4, 2003.
Brief description of amendment: The amendment changed the MSIV
full-closure stroke time of Technical Specification (TS) surveillance
requirement 4.7.1.5 from 5 seconds to 6 seconds. Additionally, the
once-per-92-day requirement to part-stroke exercise the main steam
isolation valves (MSIVs) was replaced with criteria to test each MSIV
pursuant to TS 4.0.5, which requires testing in accordance with the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code, Section XI.
Date of issuance: June 25, 2003.
Effective date: Effective the day of issuance to be implemented
within 60 days.
Amendment No: 137.
Facility Operating License No. NPF-73. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58644). The supplement dated February 4, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 25, 2003.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: July 3, 2002, as supplemented
September 24, 2002, January 10, 2003, and March 20, 2003.
Brief description of amendment: The amendment revises Improved
Technical Specification (ITS) 3.8.1 and associated Bases, ``AC Sources-
Operating,'' by extending the allowed outage time for the emergency
diesel generators from 72 hours to 14 days.
Date of issuance: June 13, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance except for installation of an Aac source. An
Aac source as described in the licensee's application supplement dated
March 20, 2003, shall be installed before completion of refueling
outage 14, as discussed in the NRC Safety Evaluation dated June 13,
2003. Implementation shall include incorporation of a description of
the Aac source into the next scheduled Final Safety Analysis Report
update after the Aac installation.
Amendment No.: 207.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50955). The September 24, 2002, January 10, 2003, and March 20, 2003,
supplements contained clarifying information only, and did not change
the initial no significant hazards consideration determination or
expand the scope of the initial application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 13, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: March 19, 2003.
Brief description of amendment: The amendment deletes Technical
Specification 6.8.C, ``Post Accident Sampling,'' and thereby eliminates
the requirements to have and maintain the post accident sampling system
at the Monticello Nuclear Generating Plant.
Date of Issuance: June 17, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 136.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25655).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 17, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear, LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: February 14, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3/4.3.4.2 to extend the surveillance test intervals
and allowed out-of-service times for the end-of-cycle recirculation
pump trip system instrumentation. In addition, the TS Bases have been
revised to address the proposed changes.
Date of issuance: June 24, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 148.
Facility Operating License No. NPF-57: This amendment revises the
TSs.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18284).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 24, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: September 26, 2002, as
supplemented on March 20, 2003.
Brief description of amendments: The amendments revise setpoint and
allowable values of the steam generator (SG) low-low level trip
function in Technical Specification (TS) Table 2.2-1, ``Reactor Trip
System Instrumentation Trip Setpoints,'' and TS Table 3.3-4,
``Engineered Safety Feature Actuation System Instrumentation Trip
Setpoints.'' The TS changes are necessary to account for a flow-induced
pressure drop through the mid-deck plate inside the SG in the SG water
level measurement.
Date of issuance: June 13, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 257 and 238.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5680). The March 20, 2003, supplement contained clarifying information
and did not change the staff's proposed finding of no significant
hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 13, 2003.
[[Page 40725]]
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M.
Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendments request: March 31, 2003, as supplemented by
letter dated April 29, 2003.
Brief description of amendments: The amendment modifies Technical
Specifications (TS) Surveillance Requirement (SR) 3.4.11.1, for Farley,
Unit 2 only by the addition of the following note that states, ``Not
required to be performed for Unit 2 for the remainder of operating
cycle 16 for Q2B31MOV8000B.'' In addition, a temporary TS SR 3.4.11.4
is added to provide compensatory action for this block valve while SR
3.4.11.1 is suspended. Further, this SR requires that power to the
Farley, Unit 2 Power Operated Relief Valve Q2B31MOV8000B be checked at
least every 24 hours for the remainder of Operating Cycle 16.
Date of issuance: June 13, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 151.
Facility Operating License No. NPF-8: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25658).
The supplement dated April 29, 2003, provided clarifying
information that did not change the scope of the March 31, 2003,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 13, 2003.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendments request: February 14, 2002, as supplemented by
letters dated July 29, 2002 and March 27, 2003.
Brief description of amendments: The amendments revise STP
technical specifications to eliminate shutdown actions associated with
radiation monitoring instrumentation. The proposed changes will enhance
plant reliability by reducing exposure to unnecessary shutdowns and
increase operational flexibility, and relax certain other restrictions.
Date of issuance: June 9, 2003.
Effective date: As of the date of issuance to be implemented within
4 months from the date of issuance.
Amendment Nos.: Unit 1--153; Unit 2--141.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15629).
The July 29, 2002, and March 27, 2003, supplemental letters
provided clarifying information that was within the scope of the
original Federal Register Notice (67 FR 15629) and did not change the
initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 9, 2003.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: December 4, 2002.
Brief description of amendments: The amendments revise several
Limiting Conditions for Operation (LCO) Notes and Required Actions in
the Technical Specifications that require suspension of operations
involving positive reactivity additions or suspension of operations
involving reactor coolant system boron concentration reductions. The
amendments revise these LCO Notes and Required Actions to allow small,
controlled, safe insertions of positive reactivity, but limit the
introduction of positive reactivity such that compliance with the
required shutdown margin or refueling boron concentration limits will
still be satisfied.
Date of issuance: June 24, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 105 and 105.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
813).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 24, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th day of June 2003.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-17028 Filed 7-7-03; 8:45 am]
BILLING CODE 7590-01-P