Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 2586-2600 [06-320]
Download as PDF
2586
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
sroberts on PROD1PC69 with NOTICES
containment pressure boundary surface
areas, a general visual-type examination,
in accordance with the Hatch 1 and 2
Qualified (N) Coatings Program, is
sufficient to inspect the subject surface
areas of the containment and will
provide an acceptable level of quality
and safety.
In summary, the licensee is proposing
an exemption from the requirements of
Section 50.55a(b)(2)(ix)(G) to use an
alternate examination method to
examine Item E.20 of Table IWE–2500–
1 of ASME Code, Section XI, pursuant
to 10 CFR 50.12(a)(1) and 10 CFR
50.12(a)(2)(ii). The licensee stated in its
application that compliance with the
visual examination requirements of
Section 50.55a(b)(2)(ix)(G) is not
necessary for accessible surface areas of
the containment vessel pressure
retaining boundary Vent System to
achieve the underlying purpose of the
rule.
3.0 Discussion
Pursuant to 10 CFR 50.12, the
Commission may, upon application by
any interested person or upon its own
initiative, grant exemptions from the
requirements of 10 CFR Part 50 when:
(1) The exemptions are authorized by
law, will not present an undue risk to
public health or safety, and are
consistent with the common defense
and security; and (2) when special
circumstances are present. Special
circumstances are present whenever, in
accordance with 10 CFR Part
50.12(a)(2)(ii), ‘‘Application of the
regulation in the particular
circumstances would not serve the
underlying purpose of the rule or is not
necessary to achieve the underlying
purpose of the rule * * *.’’ Therefore,
in determining the acceptability of the
licensee’s exemption request, the NRC
staff has performed the following
evaluation to satisfy the requirements of
10 CFR 50.12 for granting the
exemption.
The underlying purpose of 10 CFR
50.55a(b)(2)(ix)(G), as it applies to Item
E1.20 of Table IWE–2500–1, is to ensure
that an examination of the metal
containment or the metal liner of a
concrete containment is performed to
identify corrosion or other degradation
that could affect the structural or leaktight integrity of the structure.
The NRC staff examined the licensee’s
rationale to support the exemption
request and concluded that maintaining
the integrity of the coating system
applied to the Hatch 1 and 2
containment vent system components is
a preventive measure that would protect
against corrosion of the coated
components. As the licensee
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
emphasizes the effectiveness of its
coating program, the NRC staff believes
that the general visual examination
performed as part of maintaining the
integrity of the coating system is a
proactive action and will ensure the
integrity of the coated vent system
components. The proposed alternative
will provide the quality and safety level
similar to the one intended by the use
of VT–3 examination of the vent system
components, and would meet the
underlying purpose of 10 CFR Section
50.55a(b)(2)(ix)(G).
Based on a consideration of proposed
alternatives contained in the licensee’s
letters dated March 20, and August 2
and 24, 2005, the NRC staff concludes
that degradation of the containment
structure would be detected using the
proposed alternative, thus meeting the
underlying purpose of the rule.
Therefore, the NRC staff concludes that
the proposed exemption from 10 CFR
Section 50.55a(b)(2)(ix)(G) is acceptable.
4.0
Conclusion
Accordingly, the Commission has
determined that, pursuant to 10 CFR
50.12, the exemption is authorized by
law, will not present an undue risk to
the public health and safety, and is
consistent with the common defense
and security. Also, special
circumstances are present. Therefore,
the Commission hereby grants SNC an
exemption from the requirement of 10
CFR Section 50.55a(b)(2)(ix)(G) to
perform a VT–3 examination for Item
E1.2 of Table IWE–2500–1, for Hatch 1
and 2, for the 4th 10-year ISI interval.
Pursuant to 10 CFR 51.32, the
Commission has determined that the
granting of this exemption will not have
a significant effect on the quality of the
human environment (70 FR 76082).
This exemption is effective upon
issuance.
Dated at Rockville, Maryland, this 6th day
of January 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. E6–415 Filed 1–13–06; 8:45 am]
BILLING CODE 7590–01–P
PO 00000
Frm 00075
Fmt 4703
Sfmt 4703
NUCLEAR REGULATORY
COMMISSION
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses Involving No Significant
Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the
Atomic Energy Act of 1954, as amended
(the Act), the U.S. Nuclear Regulatory
Commission (the Commission or NRC
staff) is publishing this regular biweekly
notice. The Act requires the
Commission publish notice of any
amendments issued, or proposed to be
issued and 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 December
22, 2005 to January 5, 2006. The last
biweekly notice was published on
January 3, 2006 (71 FR 145).
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. Within 60 days after the
date of publication of this notice, 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
E:\FR\FM\17JAN1.SGM
17JAN1
sroberts on PROD1PC69 with NOTICES
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
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.
Normally, the Commission will not
issue the amendment until the
expiration of 60 days after the date of
publication of this notice. The
Commission may issue the license
amendment before expiration of the 60day period provided that its final
determination is that the amendment
involves no significant hazards
consideration. In addition, the
Commission may issue the amendment
prior to the expiration of the 30-day
comment period should circumstances
change during the 30-day comment
period such that failure to act in a
timely way would result, for example in
derating or shutdown of the facility.
Should the Commission take action
prior to the expiration of either the
comment period or the notice period, it
will publish in the Federal Register a
notice of issuance. Should the
Commission make a final No Significant
Hazards Consideration Determination,
any hearing will take place 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 O1F21, 11555 Rockville Pike (first
floor), Rockville, Maryland. The filing of
requests for a hearing and petitions for
leave to intervene is discussed below.
Within 60 days after the date of
publication of this notice, 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
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
consult a current copy of 10 CFR 2.309,
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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a
request for a hearing or petition for
leave to intervene is filed within 60
days, the Commission or a presiding
officer designated by the Commission or
by the Chief Administrative Judge of the
Atomic Safety and Licensing Board
Panel, will rule on the request and/or
petition; and the Secretary or the Chief
Administrative Judge of the Atomic
Safety and Licensing Board will issue a
notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, 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 general requirements: (1) The
name, address, and telephone number of
the requestor or petitioner; (2) the
nature of the requestor’s/petitioner’s
right under the Act to be made a party
to the proceeding; (3) the nature and
extent of the requestor’s/petitioner’s
property, financial, or other interest in
the proceeding; and (4) the possible
effect of any decision or order which
may be entered in the proceeding on the
requestor’s/petitioner’s interest. The
petition must also set forth the specific
contentions which the petitioner/
requestor seeks to have litigated at the
proceeding.
Each contention must consist of a
specific statement of the issue of law or
fact to be raised or controverted. In
addition, the petitioner/requestor shall
provide a brief explanation of the bases
for the contention and a concise
statement of the alleged facts or expert
opinion which support the contention
and on which the petitioner/requestor
intends to rely in proving the contention
at the hearing. The petitioner/requestor
must also provide references to those
specific sources and documents of
which the petitioner is aware and on
which the petitioner/requestor intends
to rely to establish those facts or expert
opinion. The petition must include
sufficient information to show that a
genuine dispute exists with the
applicant on a material issue of law or
PO 00000
Frm 00076
Fmt 4703
Sfmt 4703
2587
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/
requestor to relief. A petitioner/
requestor who fails to satisfy 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.
If a hearing is requested, and the
Commission has not made a final
determination on the issue of no
significant hazards consideration, 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 by:
(1) First class mail addressed to the
Office of the Secretary of the
Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001, Attention: Rulemaking and
Adjudications Staff; (2) courier, express
mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor,
One White Flint North, 11555 Rockville
Pike, Rockville, Maryland, 20852,
Attention: Rulemaking and
Adjudications Staff; (3) E-mail
addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission,
HearingDocket@nrc.gov; or (4) facsimile
transmission addressed to the Office of
the Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC,
Attention: Rulemakings and
Adjudications Staff at (301) 415–1101,
verification number is (301) 415–1966.
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 it is requested that copies be
transmitted either by means of facsimile
transmission to (301) 415–3725 or by e-
E:\FR\FM\17JAN1.SGM
17JAN1
2588
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
mail to OGCMailCenter@nrc.gov. A copy
of the request for hearing and petition
for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely requests and/or petitions
and contentions will not be entertained
absent a determination by the
Commission or the presiding officer of
the Atomic Safety and Licensing Board
that the petition, request and/or the
contentions should be granted based on
a balancing of the factors specified in 10
CFR 2.309(a)(1)(i)–(viii).
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 ADAMS Public
Electronic Reading Room on the Internet
at the NRC Web site, https://
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 PDR Reference staff at 1 (800) 397–
4209, (301) 415–4737 or by e-mail to
pdr@nrc.gov.
AmerGen Energy Company, LLC, et al.,
Docket No. 50–219, Oyster Creek
Nuclear Generating Station (OCNGS),
Ocean County, New Jersey
sroberts on PROD1PC69 with NOTICES
Date of amendment request:
December 2, 2005.
Description of amendment request:
The amendment would revise the
Technical Specifications to increase the
allowable as-found main steam safety
valve code safety function lift setpoint
tolerance from ±1% to ±3%.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Will operation of the facility in
accordance with the proposed amendment
involve a significant increase in the
probability or consequences of an accident
previously evaluated?
Response: No.
The proposed TS changes allow for an
increase in the as-found Main Steam Safety
Valve (MSSV) setpoint tolerance from ±1% to
±3%. The proposed changes do not alter the
MSSV nominal lift setpoints or MSSV lift
setpoint test frequency.
The proposed TS changes have been
evaluated on both a generic and plant
specific basis. The NRC has approved the
general approach of this change; however,
implementation is contingent on several
plant specific evaluations. The required plant
specific analyses and evaluations included
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
transient analysis of the anticipated
operational transients (AOTs); analysis of the
design basis overpressurization event;
evaluation of the performance of high
pressure systems, and evaluation of the
containment response during Loss-of-Coolant
Accident (LOCA) and hydrodynamic loads
on the MSSV discharge lines and
containment. These analyses and evaluations
demonstrate that there is adequate margin to
the design core thermal limits and reactor
vessel pressure limits using the ±3% MSSV
as-found setpoint tolerance. The analyses and
evaluations also demonstrate that the
operation of high-pressure safety systems
will not be adversely affected and that the
containment response during a LOCA will be
acceptable.
Evaluations of the impact of the proposed
change on the equipment important to safety
have been performed and no adverse
conditions were identified. The reactor
pressure vessel and attached systems and
piping have been evaluated for the impact of
this proposed TS change. A plant specific
analysis has been performed which indicates
that the ASME Code upset limits for the
reactor pressure vessel will not be exceeded
for the limiting event, i.e., Main Steam
Isolation Valve (MSIV) closure with flux
Scram. The reactor pressure vessel and
attached piping design values will not be
exceeded. Therefore, the probability of a
malfunction of the reactor pressure vessel
and attached systems and piping is not
increased and the consequences of such an
accident remain acceptable.
The nuclear fuel has been evaluated for the
impact of the proposed change.
Plant specific analyses were performed
which indicate that for all abnormal
operational transients adequate margin to the
fuel thermal limit parameters, i.e., Minimum
Critical Power Ratio (MCPR) and thermalmechanical limits, is maintained. Emergency
Core Cooling System (ECCS)/LOCA
performance is maintained adequate to meet
the requirements of 10 CFR 50.46. Therefore,
the consequences of these accidents remain
acceptable and the probability of the
malfunction of the nuclear fuel is not
increased.
The Containment response during a LOCA
has been evaluated for the impact of the
proposed change. The major factor in the
Containment pressure response to a LOCA is
the rate of reactor vessel water inventory loss
due to a DBA LOCA. The rate of reactor
vessel water inventory loss is mainly
dependent on the initial reactor pressure,
which is not affected by the proposed
setpoint tolerance change. The major factor
in the Containment temperature response to
a LOCA is the integrated steam inventory loss
due to Main Steamline Break. The rate of
reactor vessel steam inventory loss is mainly
dependent on the reactor decay heat, which
is not affected by the proposed setpoint
tolerance change. Therefore, the
consequences of these accidents remain
acceptable and the probability of the
malfunction of Containment is not increased.
The Control Rod Drive (CRD) system has
been evaluated for the impact of the
proposed change. The CRD system capability
of controlling reactor power during normal
PO 00000
Frm 00077
Fmt 4703
Sfmt 4703
plant operation and rapidly inserting control
rod blades (Scram) during abnormal plant
conditions is not impacted by the proposed
change. Therefore, the probability of a
malfunction of the CRD system is not
increased.
The Reactor Vessel Instrumentation System
has been evaluated for the impact of the
proposed change. The Reactor Vessel
Instrumentation System will continue to be
operated within the current design pressure/
temperature requirements; therefore, the
probability of a malfunction of the Reactor
Vessel Instrumentation System is not
increased.
An administrative change is also being
proposed to correct the reference to ‘‘IWV–
3510 of Section XI of the ASME Boiler and
Pressure Vessel Code’’ in TS 4.3.E because
the stated ASME section no longer exists.
The TS is being changed to reference
specification 4.3.C for MSSV testing. This is
an administrative change and does not affect
previously evaluated accidents.
Therefore, the proposed TS changes do not
significantly increase the probability or
consequences of an accident previously
evaluated.
2. Will operation of the facility in
accordance of the proposed amendment
create the possibility of a new or different
kind of accident from any accident
previously evaluated?
Response: No.
The proposed TS changes allow for an
increase in the as-found MSSV setpoint
tolerance from ±1% to ±3%. Generic and
plant specific analyses and evaluations
indicate that the plant response to any
previously evaluated event will remain
acceptable. All plant systems, structures, and
components will continue to be capable of
performing their required safety function as
required by event analysis guidance.
The proposed TS changes do not alter the
MSSV nominal lift setpoints or MSSV lift
setpoint test frequency. The operation and
response of the affected equipment important
to safety is unchanged. All systems,
structures, and components will continue to
be operated within acceptable operating and/
or design parameters. No system, structure,
or component will be subjected to a
condition that has not been evaluated and
determined to be acceptable using the
guidance required for specific event analysis.
The change to correct the reference to
‘‘IWV–3510 of Section XI of the ASME Boiler
and Pressure Vessel Code’’ in TS 4.3.E is an
administrative change and does not affect the
possibility of a new or different kind of
accident.
Therefore, the proposed TS changes do not
create the possibility of a new or different
kind of accident from any previously
identified.
3. Will operation of the facility in
accordance with the proposed amendment
involve a significant reduction in a margin of
safety?
Response: No.
The proposed TS changes allow for an
increase in the as-found MSSV setpoint
tolerance from ±1% to ±3%. The proposed
TS changes do not alter the MSSV nominal
lift setpoints or MSSV lift setpoint test
E:\FR\FM\17JAN1.SGM
17JAN1
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
frequency. The operation and response of the
affected equipment important to safety is
unchanged. All systems, structures, and
components will continue to be operated
within acceptable operating and/or design
parameters. While the calculated peak reactor
vessel pressure for the ASME overpressure
event is higher than that calculated without
the increase in setpoint tolerance, it is still
within the respective licensing acceptance
limits associated with this event. These
licensing acceptance limits have been
determined by the NRC to provide a
sufficient margin of safety.
The increase in MSSV steam flow and
reactor vessel pressure does not reduce the
margin of safety associated with the MSSVs
and associated components and structures
since the increased MSSV steam flow rate
and reactor vessel pressure are bounded by
the current design analysis.
The margin of safety for fuel thermal limits
and 10 CFR 50.46 limits are unaffected by the
proposed change.
The margin of safety for the Containment
is unaffected by the proposed change.
The capability of the SLC system and the
CRD system to perform their safety functions
during all required events, using the required
guidance for event analysis, is maintained.
Therefore, the proposed changes do not
reduce the margin of safety provided by the
SLC and CRD systems.
The change to correct the reference to
‘‘IWV–3510 of Section XI of the ASME Boiler
and Pressure Vessel Code’’ in TS 4.3.E is an
administrative change and does not affect the
margin of safety.
Therefore, these proposed TS 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: Thomas S.
O’Neill, Associate General Counsel,
Exelon Generation Company, LCC, 4300
Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Darrell J. Roberts.
sroberts on PROD1PC69 with NOTICES
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:
November 3, 2005.
Description of amendments request:
The proposed amendments would
revise the accident source term in the
design-basis radiological consequences
analyses and the associated Technical
Specifications (TSs), pursuant to section
50.67 of part 50 of Title 10 of the Code
of Federal Regulations (10 CFR 50.67).
The proposed amendments would
provide for the full implementation of
the alternate source term (AST) in
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
accordance with the guidance in
Regulatory Guide 1.183, ‘‘Alternative
Radiological Source Terms for
Evaluating Design Basis Accidents at
Nuclear Power Reactors.’’ The proposed
amendments would also increase the
flow rate for the control room
emergency ventilation system (CREVS)
from 2000 to 10000 cubic feet per
minute in TS 5.5.11, ‘‘Ventilation Filter
Testing Program,’’ by means of a
modification to the CREVS. In addition,
automatic isolation dampers and
radiation monitors will also be installed
at access control heating, ventilating,
and air conditioning (HVAC) unit no.
RTU–1 and access control air
conditioning unit no. 13.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Operation of the facility in accordance
with the proposed amendment would not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
The results of the applicable radiological
design basis accidents (DBAs) re-evaluation
demonstrated that, with the requested
changes, the dose consequences of these
limiting events are within the regulatory
limits and guidance provided by the Nuclear
Regulatory Commission in 10 CFR 50.67 and
Regulatory Guide 1.183 for AST
methodology. The AST is an input to
calculations used to evaluate the
consequences of an accident and does not by
itself affect the plant response or the actual
pathway of the activity released from the
fuel. It does, however, better represent the
physical characteristics of the release such
that appropriate mitigation techniques may
be applied.
The change from the original source term
to the new proposed AST is a change in the
analysis method and assumptions and has no
effect on accident initiators or causal factors
that contribute to the probability of
occurrence of previously analyzed accidents.
Use of an AST to analyze the dose effect of
DBAs shows that regulatory acceptance
criteria for the new methodology continues to
be met. Changing the analysis methodology
does not change the sequence or progression
of the accident scenario.
The proposed Technical Specification
changes reflect the plant configuration that
will either support implementation of the
AST analyses or eliminate requirements that
are no longer needed as a result of the revised
DBA analyses. The equipment affected by the
proposed changes is mitigative in nature and
relied upon after an accident has been
initiated. The operation of various filtration
systems have been considered in the
evaluations for these proposed changes.
While the operation of some systems does
change with the implementation of an AST,
the affected systems are not accident
PO 00000
Frm 00078
Fmt 4703
Sfmt 4703
2589
initiators; and application of the AST
methodology, itself, is not an initiator of a
DBA.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Operation of the facility in accordance
with the proposed amendment would not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
As described in Item 1 above, the changes
proposed in this license amendment request
involve the use of a new analysis
methodology and related regulatory
acceptance criteria. The proposed Technical
Specification changes reflect the plant
configuration that will either support
implementation of the new methodology or
eliminate requirements that are no longer
needed as a result of the new methodology.
No new or different accidents result from
utilizing the proposed changes. Although the
proposed changes require modification to the
Control Room emergency ventilation system
and installation of automatic isolation
dampers and radiation monitors at Access
Control HVAC Unit RTU–1 and Access
Control Air Conditioning Unit 13 on the
Auxiliary Building roof, none of these
changes can initiate a new or different kind
of accident since they are only related to
system capabilities that provide protection
from accidents that have already occurred.
As a result, no new failure modes are being
introduced that could lead to different
accidents. These changes do not alter the
nature of events postulated in the Updated
Final Safety Analysis Report nor do they
introduce any unique precursor mechanisms.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any previously
evaluated.
3. Operation of the facility in accordance
with the proposed amendment would not
involve a significant reduction in a margin of
safety.
As described in Item 1 above, the changes
proposed in this license amendment request
involve the use of a new analysis
methodology and related regulatory
acceptance criteria. The proposed Technical
Specification changes reflect the plant
configuration that will either support
implementation of the new methodology or
eliminate requirements that are no longer
needed as a result of the new methodology.
Safety margins and analytical conservatisms
have been evaluated and have been found
acceptable. The analyzed events have been
carefully selected and, with plant
modification, margin has been retained to
ensure that the analyses adequately bound
postulated event scenarios. The analyses
have been performed using conservative
methodologies, as specified in Regulatory
Guide 1.183. The dose consequences of these
DBAs remain within the acceptance criteria
presented in 10 CFR 50.67, ‘‘Accident Source
Term,’’ and Regulatory Guide 1.183. The
proposed changes continue to ensure that the
doses at the exclusion area boundary and low
population zone boundary, as well as the
Control Room, are within corresponding
regulatory limits.
E:\FR\FM\17JAN1.SGM
17JAN1
2590
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
sroberts on PROD1PC69 with NOTICES
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 amendments
request involves no significant hazards
consideration.
Attorney for licensee: Carey Fleming, Sr.
Counsel—Nuclear Generation, Constellation
Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket
No. 50–352, Limerick Generating Station,
Unit 1, Montgomery County, Pennsylvania
Date of amendment request: December 14,
2005.
Description of amendment request: The
proposed amendment modifies the Technical
Specifications (TSs) to incorporate a revised
Single Loop Operation Safety Limit
Minimum Critical Power Ratio (SLO
SLMCPR) due to the cycle-specific analysis.
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 derivation of the cycle specific Single
Loop Operation Safety Limit Minimum
Critical Power Ratio (SLO SLMCPR) for
incorporation into the Technical
Specifications (TS), and its use to determine
cycle-specific thermal limits, has been
performed using the methodology discussed
in ‘‘General Electric Standard Application for
Reactor Fuel,’’ NEDE–24011–P–A–15
(GESTAR–II), and U.S. Supplement, NEDE–
24011–P–A–15–US, September, 2005, which
includes Amendment 25. Amendment 25
was approved by the NRC in a March 11,
1999 safety evaluation report.
The basis of the SLO SLMCPR calculation
is to ensure that greater than 99.9% of all fuel
rods in the core avoid transition boiling if the
limit is not violated. The new SLO SLMCPR
preserves the existing margin to transition
boiling. The GE–14 fuel is in compliance
with Amendment 22 to ‘‘General Electric
Standard Application for Reactor Fuel,’’
NEDE–24011–P–A–15 (GESTAR–II), and U.S.
Supplement, NEDE–24011–P–A–15–US,
September 2005, which provides the fuel
licensing acceptance criteria. The probability
of fuel damage will not be increased as a
result of this change. Therefore, the proposed
TS 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.
The SLO SLMCPR is a TS numerical value,
calculated to ensure that transition boiling
does not occur in 99.9% of all fuel rods in
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
the core if the limit is not violated. The new
SLO SLMCPR is calculated using NRC
approved methodology discussed in ‘‘General
Electric Standard Application for Reactor
Fuel,’’ NEDE–24011–P–A–15 (GESTAR–II),
and U.S. Supplement, NEDE–24011–P–A–
15–US, September 2005, which includes
Amendment 25. Additionally, the GE–14 fuel
is in compliance with Amendment 22 to
‘‘General Electric Standard Application for
Reactor Fuel,’’ NEDE–24011–P–A–15
(GESTAR–II), and U.S. Supplement, NEDE–
24011–P–A–15–US, September, 2005, which
provides the fuel licensing acceptance
criteria. The SLO SLMCPR is not an accident
initiator, and its revision will not create the
possibility of a new or different kind of
accident from any accident previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
There is no significant reduction in the
margin of safety previously approved by the
NRC as a result of the proposed change to the
SLO SLMCPR, which includes the use of GE–
14 fuel. The new SLO SLMCPR is calculated
using methodology discussed in ‘‘General
Electric Standard Application for Reactor
Fuel,’’ NEDE–24011–P–A–15 (GESTAR–II),
and U.S. Supplement, NEDE–24011–P–A–
15–US, September, 2005, which includes
Amendment 25. The SLO SLMCPR ensures
that greater than 99.9% of all fuel rods in the
core will avoid transition boiling if the limit
is not violated when all uncertainties are
considered, thereby preserving the fuel
cladding integrity.
Therefore, the proposed TS change will not
involve a significant reduction in [a] margin
of safety previously approved by the NRC.
The NRC staff has reviewed the licensee’s
analysis and, based on this review, it appears
that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff
proposes to determine that the amendment
request involves no significant hazards
consideration.
Attorney for licensee: Mr. Brad Fewell,
Assistant General Counsel, Exelon
Generation Company, LLC, 200 Exelon Way,
Kennett Square, PA 19348.
NRC Branch Chief: Darrell J. Roberts.
Exelon Generation Company, LLC, Docket
Nos. 50–352 and 50–353, Limerick
Generating Station, Units 1 and 2,
Montgomery County, Pennsylvania
Date of amendment request: December 21,
2005.
Description of amendment request: The
proposed amendment revises the Technical
Specifications by relocating the Pressure
Isolation Valve (PIV) tables to the Technical
Requirements Manual (TRM).
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 relocation of Technical
Specification Table 3.4.3.2–1 does not alter
PO 00000
Frm 00079
Fmt 4703
Sfmt 4703
the requirements for pressure isolation valve
operability or surveillance currently in the
Technical Specifications. The proposed
change to remove the pressure isolation valve
table from TS and relocate the information to
an administratively controlled document,
and to revise the wording in TS to reflect this
change, will have no impact on any safety
related structures, systems or components.
The probability of occurrence of a previously
evaluated accident is not increased because
this change does not introduce any new
potential accident initiating conditions. The
consequences of accidents previously
evaluated in the UFSAR [Updated Final
Safety Analysis Report] are not affected
because the ability of the PIVs to limit
leakage through these valves in amounts that
do not compromise safety is not affected.
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 changes are administrative
in nature and do not result in physical
alterations or changes in the method by
which any safety related system performs its
intended function(s). The proposed changes
do not impact any safety analysis
assumptions. The proposed changes do not
create any new accident initiators or involve
an activity that could be an initiator of an
accident of a different type.
All PIVs and alarm instrumentation will
continue to be tested to the same rigorous
requirements as defined in the Technical
Specification Surveillance Requirements.
The proposed revision does not make
changes in any method of testing or how any
safety related system performs its safety
functions. Therefore, the possibility of an
accident of a different type than any
previously evaluated in the UFSAR is not
created.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The administrative change to relocate
Technical Specification Table 3.4.3.2–1 to
the Technical Requirements Manual does not
alter the basic regulatory requirement for
Reactor Coolant System pressure isolation
and will not affect the isolation capability for
credible accident scenarios. Future revisions
to the Technical Requirements Manual Table
will be subject to evaluation pursuant to 10
CFR 50.59.
Additionally, the proposed relocation does
not alter the requirements for pressure
isolation valve and alarm instrumentation
operability currently in the Technical
Specifications. The LCO [limiting condition
for operation] and Surveillance Requirements
will be retained in the revised Technical
Specifications. The proposed change will not
affect the meaning, application, and function
of the current Technical Specification
requirements for the valves in Table 3.4.3.2–
1. Therefore, the proposed changes do not
result in a significant reduction in [a] margin
of safety.
E:\FR\FM\17JAN1.SGM
17JAN1
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. Brad
Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200
Exelon Way, Kennett Square, PA 19348.
NRC Branch Chief: Darrell J. Roberts.
sroberts on PROD1PC69 with NOTICES
Exelon Generation Company, LLC
(EGC, licensee), Docket No. 50–265,
Quad Cities Nuclear Power Station
(QCNPS), Unit 2, Rock Island County,
Illinois
Date of amendment request:
December 15, 2005.
Description of amendment request:
The proposed change revises the values
of the safety limit minimum critical
power ratio (SLMCPR) in Technical
Specification (TS) section 2.1.1,
‘‘Reactor Core SLs.’’ Specifically, the
proposed change would require that for
Unit 2, the minimum critical power
ratio (MCPR) for Global Nuclear Fuel
(GNF) fuel shall be ≥1.09 for two
recirculation loop operation, or ≥1.10
for single recirculation loop operation.
Additionally, the proposed change
would require that MCPR for
Westinghouse fuel shall be ≥1.11 for two
recirculation loop operation, or ≥1.13
for single recirculation loop 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:
According to 10 CFR 50.92, ‘‘Issuance of
amendment,’’ paragraph (c), a proposed
amendment to an operating license involves
no significant hazards consideration if
operation of the facility in accordance with
the proposed amendment would not:
(1) Involve a significant increase in the
probability or consequences of an accident
previously evaluated; or
(2) Create the possibility of a new or
different kind of accident from any accident
previously evaluated; or
(3) Involve a significant reduction in a
margin of safety.
EGC has evaluated the proposed change to
the TS for QCNPS, Unit 2, using the criteria
in 10 CFR 50.92, and has determined that the
proposed change does not involve a
significant hazards consideration. The
following information is provided to support
a finding of no significant hazards
consideration.
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
Response: No.
The probability of an evaluated accident is
derived from the probabilities of the
individual precursors to that accident. The
consequences of an evaluated accident are
determined by the operability of plant
systems designed to mitigate those
consequences. Limits have been established
consistent with NRC-approved methods to
ensure that fuel performance during normal,
transient, and accident conditions is
acceptable. The proposed change
conservatively establishes the SLMCPR for
QCNPS, Unit 2, Cycle 19 such that the fuel
is protected during normal operation and
during plant transients or anticipated
operational occurrences (AOOs).
Changing the SLMCPR does not increase
the probability of an evaluated accident. The
change does not require any physical plant
modifications, physically affect any plant
components, or entail changes in plant
operation. Therefore, no individual
precursors of an accident are affected.
The proposed change revises the SLMCPR
to protect the fuel during normal operation
as well as during plant transients or AOOs.
Operational limits will be established based
on the proposed SLMCPR to ensure that the
SLMCPR is not violated. This will ensure
that the fuel design safety criterion (i.e., that
at least 99.9% of the fuel rods do not
experience transition boiling during normal
operation and AOOs) is met. Since the
proposed change does not affect operability
of plant systems designed to mitigate any
consequences of accidents, the consequences
of an accident previously evaluated are not
expected to increase.
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.
Creation of the possibility of a new or
different kind of accident would require
creating one or more new accident
precursors. New accident precursors may be
created by modifications of plant
configuration, including changes in
allowable modes of operation.
The proposed change does not involve any
plant configuration modifications or changes
to allowable modes of operation. The
proposed change to the SLMCPR assures that
safety criteria are maintained for QCNPS,
Unit 2, Cycle 19.
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 SLMCPR provides a margin of safety
by ensuring that at least 99.9% of the fuel
rods do not experience transition boiling
during normal operation and AOOs if the
MCPR limit is not violated. The proposed
change will ensure the appropriate level of
fuel protection by continuing to ensure that
at least 99.9% of the fuel rods do not
PO 00000
Frm 00080
Fmt 4703
Sfmt 4703
2591
experience transition boiling during normal
operation and AOOs if the MCPR limit is not
violated. Additionally, operational limits will
be established based on the proposed
SLMCPR to ensure that the SLMCPR is not
violated. This will ensure that the fuel design
safety criteria (i.e., that no more than 0.1%
of the rods are expected to be in boiling
transition if the MCPR limit is not violated)
are met.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
Based upon the above, EGC concludes that
the proposed amendment presents no
significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and,
accordingly, a finding of no significant
hazards consideration is justified.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
requested amendments involve no
significant hazards consideration.
Attorney for licensee: Mr. Brad
Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 4300
Winfield Road, Warrenville, IL 60555.
NRC Acting Branch Chief: Mindy S.
Landau.
First Energy Nuclear Operating
Company, Docket No. 50–440, Perry
Nuclear Power Plant, Unit 1 (PNPP),
Lake County, Ohio
Date of amendment request:
November 21, 2005.
Description of amendment request:
The proposed amendment would revise
the acceptance criteria of Technical
Specification (TS) Surveillance
Requirements (SRs) associated with TS
3.8.1, ‘‘AC Sources—Operating,’’ to
modify the Emergency Diesel Generator
(EDG) start tests to provide minimum
voltage and frequency limits and clarify
other limits as steady state parameters.
Specifically, the amendment would
revise SRs 3.8.1.2, 3.8.1.7, 3.8.1.12,
3.8.1.15 and 3.8.1.20. This change is
consistent with the approved Technical
Specification Task Force Traveler
(TSTF) 163, Revision 2.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve
a significant increase in the probability or
consequences of an accident previously
evaluated.
The proposed change is a LAR (license
amendment request) that modifies the
acceptance criteria for the PNPP TS SRs
pertaining to the EDGs. The EDGs mitigate
the consequences of previously evaluated
E:\FR\FM\17JAN1.SGM
17JAN1
sroberts on PROD1PC69 with NOTICES
2592
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
accidents involving a loss of offsite power.
The EDGs are used to support mitigation of
the consequences of an accident, but they are
not considered as the initiator of any
previously analyzed accident.
The proposed LAR does not change the
manner in which the EDGs are operated and
when implemented will continue to ensure
the EDGs perform their function when called
upon. The proposed revision to the TS SRs
will continue to ensure that minimum
frequency and voltage are attained within the
required time. The SRs will continue to
ensure that proper steady state voltage and
frequency are attained consistent with proper
EDG governor and voltage regulator
performance.
The proposed LAR does not affect the
design of the EDGs, the operational
characteristics of the EDGs, the interfaces
between the EDGs and other plant systems,
the function, or reliability of the EDGs. Thus,
the EDGs will be capable of performing their
accident mitigation function and there is no
impact to the radiological consequences of
any accident analysis.
As such, the proposed change continues to
provide adequate assurance of operable EDGs
and does not involve any increase to the
probability or consequences of an accident
previously evaluated.
2. The proposed change would not create
the possibility of a new or different kind of
accident from any previously evaluated.
The proposed LAR introduces no new
mode of plant operation and it does not
involve physical modification to the plant.
New equipment is not installed with the
proposed LAR, nor does the proposed LAR
cause existing equipment to be operated in a
new or different manner.
Since the proposed changes do not involve
a change to the plant design or operation, no
new system interactions are created by this
change. The proposed LAR does not produce
any parameters or conditions that could
contribute to the initiation of accidents
different from those already evaluated in the
Updated Safety Analysis Report.
The changes to the affected TS SRs do not
affect the assumed accident performance of
the EDGs, nor any plant structure, system or
component previously evaluated.
Therefore, the proposed LAR does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. The proposed change will not involve a
significant reduction in the margin of safety.
The proposed change is a LAR that does
not impact EDG performance, including the
capability for each EDG to attain and
maintain required voltage and frequency for
accepting and supporting plant safety loads
within the required time, as assumed in the
plant safety analysis.
The proposed LAR does not involve a
significant reduction in a margin of safety
since the operability of the EDGs continues
to be determined as required to support the
capability of the EDGs to provide emergency
power to plant equipment that mitigate the
consequences of an accident.
The proposed LAR does not introduce
changes to setpoints or limits established or
assumed by the accident analysis. Therefore,
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
implementation of the proposed LAR 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: David W.
Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: Mindy Landau,
Acting.
FPL Energy Seabrook, LLC, Docket No.
50–443, Seabrook Station, Unit No. 1,
Rockingham County, New Hampshire
Date of amendment request:
December 6, 2005.
Description of amendment request:
The proposed amendment would revise
the Seabrook Station, Unit No. 1
Technical Specification 3.8.3.1, ‘‘Onsite
Power Distribution,’’ to extend the
allowed outage time for balance-of-plant
vital inverters 1–EDE–I–1E and 1–EDE–
I–1F from 24 hours to 7 days.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. The proposed changes do not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
The proposed change extends the allowed
outage time (AOT) for the balance-of-plant
(BOP) instrument bus inverters from 24 hours
to 7 days. The BOP instrument bus inverters
do not solely support any risk-significant
functions. The failure of an inverter is not an
initiator of any analyzed event and does not
increase the frequency of an initiating event.
Consequently, extending the AOT will not
have an impact on the frequency of
occurrence of any event previously analyzed.
The proposed change does not alter the
design, configuration, operation, or function
of any plant system, structure, or component.
As a result, the outcomes of previously
evaluated accidents are unaffected.
Therefore, the proposed change does 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 kind of
accident from any accident previously
evaluated.
No new accident scenarios, failure
mechanisms, or limiting single failures are
introduced as a result of the proposed
change. The proposed change does not
challenge the performance or integrity of any
safety-related system. The proposed change
PO 00000
Frm 00081
Fmt 4703
Sfmt 4703
neither installs nor removes any plant
equipment, not alters the design, physical
configuration, or mode of operation of any
plant structure, system, or component.
Installed equipment will not be operated in
a new or different manner. No physical
changes are being made to the plant, so no
new accident causal mechanisms are being
introduced. Procedures that ensure the unit
operates within analyzed limits and
procedures that respond to off-normal and
emergency conditions are not altered with
this proposed change. Therefore, the
proposed change does not create the
possibility of a new or different accident
from any previously evaluated.
3. The proposed changes do not involve a
significant reduction in [a] margin of safety.
The margin of safety associated with the
acceptance criteria of any accident is
unchanged. The proposed change does not
alter the design, configuration, operation, or
function of any plant system, structure, or
component. The ability of any operable
structure, system, or component to perform
its designated safety function is unaffected by
this change. Operation with one instrument
bus inverter inoperable and the associated
instrument bus aligned to its maintenance
supply does not result in a significant
reduction in [a] margin of safety.
Surveillance testing of the emergency diesel
generators (EDGs) and the electrical
distribution system provides confidence that
the EDGs will energize the emergency AC
buses following a loss of power. 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: M. S. Ross,
Florida Power & Light Company, P.O.
Box 14000, Juno Beach, FL 33408–0420.
NRC Branch Chief: Darrell J. Roberts.
Nuclear Management Company, LLC,
Docket Nos. 50–266 and 50–301, Point
Beach Nuclear Plant, Units 1 and 2,
Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request:
November 12, 2004.
Description of amendment request:
The proposed amendment would revise
Technical Specification (TS) 5.5.7,
‘‘Inservice Testing Program,’’ and TS
5.5.8, ‘‘Steam Generator (SG) Tube
Surveillance Program,’’ to update
references to the American Society of
Mechanical Engineers (ASME) Boiler
and Pressure Vessel Code (Code) and
certain associated periodicities for
inservice testing activities consistent
with the requirements of Title 10 of the
Code of Federal Regulations (10 CFR)
section 50.55a, ‘‘Codes and standards.’’
E:\FR\FM\17JAN1.SGM
17JAN1
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
sroberts on PROD1PC69 with NOTICES
The proposed amendment would also
correct a typographical error contained
in TS 5.5.8.b.2.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear
Plant in accordance with the proposed
amendments does not result in a significant
increase in the probability or consequences
of any accident previously evaluated.
The proposed change revises Technical
Specifications for consistency with the
requirements of 10 CFR 50.55a(f)(4) and 10
CFR 50.55a(g)(4).
The proposed change incorporates
revisions to the ASME Code that result in a
net improvement in the measures for testing
pumps and valves.
The proposed 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
actuation setpoints, accident mitigation
capabilities, or accident analysis assumptions
or inputs.
Therefore, the probability or consequences
of any accident previously evaluated will not
be significantly increased as a result of the
proposed change.
2. Operation of the Point Beach Nuclear
Plant in accordance with the proposed
amendments does not result in a new or
different kind of accident from any accident
previously evaluated.
The proposed change incorporates
revisions to the ASME Code that result in a
net improvement in the measures for testing.
The proposed change does not involve a
modification to the physical configuration of
the plant (i.e., no new equipment will be
installed) or change in the methods
governing normal plant operation. The
proposed change will not impose any new or
different requirements or introduce a new
accident initiator, accident precursor, or
malfunction mechanism. Additionally, there
is no change in the types or increases in the
amounts of any effluent that may be released
off-site and there is no increase in individual
or cumulative occupational exposure.
Equipment important to safety will
continue to operate as designed. The changes
do not result in any event previously deemed
incredible been made credible. The changes
do not result in adverse conditions or result
in any increase in the challenges to safety
systems. Therefore, operation of the Point
Beach Nuclear Plant in accordance with the
proposed amendment will not create the
possibility of a new or different type of
accident from any accident previously
evaluated.
3. Operation of the Point Beach Nuclear
Plant in accordance with the proposed
amendments does not result in a significant
reduction in a margin of safety.
The proposed change incorporates
revisions to the ASME Code that result in a
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
net improvement in the measures for testing.
The safety function of the affected
components will be maintained.
There are no new or significant changes to
the initial conditions contributing to accident
severity or consequences. The proposed
amendment will not otherwise affect the
plant protective boundaries, will not cause a
release of fission products to the public, nor
will it degrade the performance of any other
structures, systems or components (SSCs)
important to safety. Therefore, the requested
change will not result in 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: Jonathan Rogoff,
Esquire, Vice President, Counsel &
Secretary, Nuclear Management
Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50–354,
Hope Creek Generating Station, Salem
County, New Jersey
Date of amendment request: October
11, 2005.
Description of amendment request:
The proposed amendment would revise
certain 18-month Technical
Specification (TS) Surveillance
Requirements (SRs) to eliminate the
condition that testing be conducted
during shutdown.
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 changes permit PSEG to
evaluate the conditions required to safely
perform a TS SR. These surveillance tests
verify that equipment will perform its
intended safety function of mitigating an
accident. No analyzed accident scenario is
being revised. The initiating conditions and
assumptions for accidents described in the
Hope Creek Generating Station Updated
Final Safety Analysis Report (UFSAR) remain
as previously analyzed.
The proposed changes do not reduce the
ability of the mitigating equipment to
perform its safety function. The TS will
continue to require the surveillance tests to
be performed on an eighteen-month
periodicity to verify operability. As a result,
the ability of the mitigating equipment to
PO 00000
Frm 00082
Fmt 4703
Sfmt 4703
2593
perform its safety function is unaffected by
the proposed change.
The capitalization change is proposed to
improve readability and does not alter any
requirement.
Based upon the above, the proposed
changes will not involve a significant
increase in the probability or consequences
of an accident previously analyzed.
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 create the
possibility of a new or different kind of
accident from any accident previously
evaluated in the UFSAR. No new accident
scenarios, failure mechanisms, or limiting
single failures are introduced as a result of
the proposed changes. Specifically, no new
hardware is being added to the plant as part
of the proposed change, no existing
equipment is being modified, and no
significant changes in operations are being
introduced.
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 will not alter any
assumptions, initial conditions, or results of
any accident analyses. The proposed changes
to remove the requirement to perform certain
testing during shutdown conditions allows
PSEG to evaluate the conditions needed to
safely perform the required testing. There is
no change to the frequency of testing or in
the testing that is required. There is no
change in the responsibility of PSEG to
perform tests in a safe and responsible
manner. Any changes to procedures will
have to be individually evaluated to ensure
that they do not reduce the margin of safety.
The changes do not affect the ability of
systems, structures or components to perform
their safety related functions. In addition, the
proposed changes do not affect the ability of
the safety systems to ensure that the facility
can be maintained in a shutdown or refueling
condition for extended periods of time.
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 Branch Chief: Darrell J. Roberts.
E:\FR\FM\17JAN1.SGM
17JAN1
2594
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
PSEG Nuclear LLC, Docket Nos. 50–272
and 50–311, Salem Nuclear Generating
Station, Unit Nos. 1 and 2, Salem
County, New Jersey
sroberts on PROD1PC69 with NOTICES
Date of amendment request: August
31, 2005; as supplemented December 8,
2005.
Description of amendment request:
The proposed amendment would
relocate the containment high range
accident monitors from the radiation
monitoring instrumentation technical
specification (TS) to the accident
monitoring TS and correct a
typographical error contained in a
previous amendment.
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 presents no change
in the probability of a previously evaluated
accident.
The proposed change presents no change
in the consequence of an accident, since the
containment high range accident monitors
are used post-accident to determine the
amount of core damage and status of the
fission product barriers.
The containment high range accident
monitors are used post accident to assess the
conditions inside containment. They have an
automatic function to switch the subcooling
margin monitor (SCMM) to ‘‘adverse’’ mode
(i.e., it displays a more conservative
indication of the amount of subcooling in the
RCS) [reactor coolant system]. Additionally,
the containment high range accident
monitors provide an indication that is used
post accident in determining the status of the
fission product barriers. There will be no
change in the operation or use of the
containment high range accident monitors.
The remaining change is editorial in nature
and does not impact the accident analysis in
any manner.
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 is a minor change
that is administrative in nature. No new
accident scenarios, failure mechanisms, or
limiting single failures are introduced as a
result of the proposed changes. No new
hardware is added, existing hardware is not
modified and no significant changes in
operations are implemented. Post accident
monitoring instrumentation is not associated
with the initiation of an accident.
3. Does the proposed change involve a
significant reduction in [a] margin of safety?
Response: No.
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
The proposed change does not alter the
manner in which safety limits, limiting safety
systems settings or limiting conditions for
operation are determined. The proposed
change will not alter any assumptions, initial
conditions or results specified in any
accident analysis.
There is no change in the containment high
range accident monitor high level alarm
setpoint. The ECS [electronic check source]
is functionally equivalent to the TS
definition of SOURCE CHECK.
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 Branch Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket No. 50–311,
Salem Nuclear Generating Station, Unit
No. 2, Salem County, New Jersey
Date of amendment request:
September 21, 2005.
Description of amendment request:
The amendment would change the
scope of steam generator (SG) tube
inspections required in the SG tubesheet
region.
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.
Of the various accidents previously
evaluated, the proposed changes only affect
the steam generator tube rupture (SGTR)
event evaluation and the postulated steam
line break (SLB) accident evaluation. Loss-ofcoolant accident (LOCA) conditions cause a
compressive axial load to act on the tube.
Therefore, since the LOCA tends to force the
tube into the tubesheet rather than pull it out,
it is not a factor in this amendment request.
Another faulted load consideration is a safe
shutdown earthquake (SSE); however, the
seismic analysis of Westinghouse 51 Series
SGs has shown that axial loading of the tubes
is negligible during an SSE.
PSEG’s amendment request takes credit for
how the tubesheet enhances the tube
integrity in the Westinghouse Electric
Company explosive tube expansion
(WEXTEX) region by precluding tube
deformation beyond its initial expanded
outside diameter. For the SGTR and SLB
events, the required structural margins of the
SG tubes will be maintained due to the
PO 00000
Frm 00083
Fmt 4703
Sfmt 4703
presence of the tubesheet. Tube rupture is
precluded for axial cracks in the WEXTEX
region due to the constraint provided by the
tubesheet. Therefore, the normal operating
3DP margin and the postulated accident
1.43DP margin against burst are maintained.
The W* length supplies the necessary
resistive force to preclude pullout loads
under both normal operating and accident
conditions. The contact pressure results from
the WEXTEX expansion process, thermal
expansion mismatch between the tube and
tubesheet, and from the differential pressure
between the primary and secondary side.
Therefore, the proposed change results in no
significant increase in the probability or the
occurrence of an SGTR or SLB accident.
The proposed changes do not affect other
systems, structures, components or
operational features. Therefore, based on the
above evaluation, the proposed changes do
not involve a significant increase in the
probability of an accident previously
evaluated.
The consequences of an SGTR event are
primarily affected by the primary-tosecondary flow rate and the time duration of
the primary-to-secondary flow during the
event. Primary-to-secondary flow rate
through a postulated ruptured tube (i.e.,
complete severance of a single SG tube) is not
affected by the proposed change since the
flow rate is based on the inside diameter of
a[n] SG tube and the pressure differential.
PSEG’s amendment request does not change
either of these. The duration of primary-tosecondary leakage is based on the time
required for an operator to determine that
a[n] SGTR has occurred, the time to identify
and isolate the faulted SG, and ensure
termination of radioactive release to the
atmosphere from the faulted SG. PSEG’s
amendment request does not affect the
duration of the primary-to-secondary leakage
because it does not change the control room
indicators with which an operator would
determine that an SGTR has occurred. The
consequences of an SGTR are secondarily
affected by primary-to-secondary leakage,
which could occur due to axial cracks
remaining in service in the WEXTEX region
in a non-faulted SG. During a[n] SGTR, the
primary-to-secondary differential pressure is
less than or equal to the normal operating
differential pressure; therefore, the primaryto-secondary leakage due to axial cracks in
the WEXTEX region of a non-faulted SG
during a[n] SGTR would be less than or equal
to the primary-to-secondary leakage
experienced during normal operation.
Primary-to-secondary leakage is considered
in the calculation determining the
consequences of a[n] SGTR and the value is
bounding.
The postulated SLB has the greatest
primary-to-secondary pressure differential,
and therefore could experience the greatest
primary-to-secondary leakage. PSEG’s
amendment request requires the aggregate
leakage, (i.e., the combined leakage for the
tubes with service induced degradation
inside the tubesheet) to remain below the
maximum allowable SLB primary-tosecondary leakage rate limit such that the
doses are maintained to less than the 10 CFR
[Part] 100 limits and also less than the GDC[General Design Criterion]19 limits.
E:\FR\FM\17JAN1.SGM
17JAN1
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
sroberts on PROD1PC69 with NOTICES
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.
PSEG’s amendment request does not
introduce any physical changes to the Salem
Unit 2 SGs. PSEG’s amendment request takes
credit for how the tubesheet enhances the SG
tube integrity in the WEXTEX region.
Because degradation detected within the W*
distance are required to be plugged, it is
highly unlikely that a tube would fail as a
result of a circumferential defect. Therefore
a tube severance, which would strike
neighboring tubes and create a multiple tube
rupture, is not credible. The proposed change
does not introduce any new equipment or
any change to existing equipment. No new
effects on existing equipment are created.
Based on the above evaluation, the proposed
changes do not create the possibility of a new
or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The amendment request maintains the
structural margins of the SG tubes for both
normal and accident conditions that are
required by Regulatory Guide 1.121. For
cracking located within the tubesheet, tube
burst is precluded due to the presence of the
tubesheet. WCAP–14797, Revision 2 defines
a length W* of degradation free expanded
tubing, that provides the necessary resistance
to tube pullout due to the pressure induced
forces (with applicable safety factor applied).
Application of the W* methodology will
preclude unacceptable primary-to-secondary
leakage during all plant conditions. The
methodology for determining leakage
provides for large margins between
calculated and actual leakage values in the
W* criteria.
Based on the above, it is concluded that the
proposed changes do not result in a
significant reduction of margin with respect
to plant safety as defined in the Updated
Final Analysis Report or Technical
Specifications. 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: Jeffrie J. Keenan,
Esquire, Nuclear Business Unit—N21,
P.O. Box 236, Hancocks Bridge, NJ
08038.
NRC Branch Chief: Darrell J. Roberts.
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
PPL Susquehanna, LLC, Docket No. 50–
387, Susquehanna Steam Electric
Station, Unit 1 (SSES 1), Luzerne
County, Pennsylvania
Date of amendment request:
December 1, 2005.
Description of amendment request:
The proposed amendment would
change the SSES–1 Technical
Specifications (TSs) by revising the Unit
1 Cycle 15 (U1C15) minimum critical
power ratio (MCPR) safety limit for
single loop operation in section 2.1.1.2
and references listed in TS 5.6.5.b.
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 of
occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change to the single-loop
MCPR Safety Limit does not directly or
indirectly affect any plant system,
equipment, component, or change the
processes used to operate the plant. Further,
the proposed U1C15 MCPR Safety Limit was
generated using NRC approved methodology
and meets the applicable acceptance criteria.
Thus, this proposed amendment does not
involve a significant increase in the
probability of occurrence or consequences of
an accident previously evaluated.
Prior to the startup of U1C15, licensing
analyses are performed (using NRC approved
methodology referenced in Technical
Specification Section 5.6.5.b) to determine
changes in the critical power ratio as a result
of anticipated operational occurrences. These
results are added to the MCPR Safety Limit
values to generate the MCPR operating limits
in the U1C15 COLR [core operating limits
report]. These limits could be different from
those specified for the current Unit 1 COLR.
The COLR operating limits thus assure that
the MCPR Safety Limit will not be exceeded
during normal operation or anticipated
operational occurrences. Postulated accidents
are also analyzed prior to the startup of
U1C15 and the results shown to be within
the NRC approved criteria.
The changes to the references in Section
5.6.5.b were made to properly reflect the NRC
approved methodology used to generate the
U1C15 core operating limits. The use of this
approved methodology does not increase the
probability of occurrence or consequences of
an accident previously evaluated.
Therefore, the proposed amendment does
not involve a significant increase in the
probability of occurrence 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 change to the single-loop MCPR Safety
Limit does not directly or indirectly affect
PO 00000
Frm 00084
Fmt 4703
Sfmt 4703
2595
any plant system, equipment, or component
and therefore does not affect the failure
modes of any of these items. Thus, the
proposed change does not create the
possibility of a previously unevaluated
operator error or a new single failure. The
changes to the references in Section 5.6.5.b
were made to properly reflect the NRC
approved methodology used to generate the
U1C15 core operating limits. The use of this
approved methodology does not create the
possibility of a new or different kind of
accident.
Therefore, this proposed amendment 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.
Since the proposed changes do not alter
any plant system, equipment, component, or
the processes used to operate the plant, the
proposed change will not jeopardize or
degrade the function or operation of any
plant system or component governed by
Technical Specifications. The proposed
single-loop MCPR Safety Limit does not
involve a significant reduction in the margin
of safety as currently defined in the Bases of
the applicable Technical Specification
sections, because the MCPR Safety Limits
calculated for U1C15 preserve the required
margin of safety.
The changes to the references in section
5.6.5.b were made to properly reflect the NRC
approved methodology used to generate the
U1C15 core operating limits. This approved
methodology is used to demonstrate that all
applicable criteria are met, thus,
demonstrating that there is no reduction in
the margin of safety.
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: Bryan A. Snapp,
Esquire, Assoc. General Counsel, PPL
Services Corporation, 2 North Ninth St.,
GENTW3, Allentown, PA 18101–1179.
NRC Branch Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos.
50–387 and 50–388, Susquehanna
Steam Electric Station, Units 1 and 2
(SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: October
5, 2005.
Description of amendment request:
The proposed amendment would revise
the SSES 1 and 2 Technical
Specifications (TSs) 3.4.10, ‘‘RCS
[reactor coolant system] Pressure and
Temperature (P/T) Limits,’’ to remove
valid P/T curve limit date and replacing
E:\FR\FM\17JAN1.SGM
17JAN1
2596
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
sroberts on PROD1PC69 with NOTICES
it with the effective full-power years
(EFPY) of radiation exposure on each of
the P/T limit curves for SSES 1 and 2.
The new P/T limit would be 35.7 EFPY
for SSES 1 and 30.2 EFPY for SSES 2.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
No. The proposed changes request that the
P/T limits curves in TS 3.4.10, ‘‘RCS Pressure
and Temperature (P/T) Limits’’ be revised by
removing the valid date and replacing it with
the Effective Full Power Years of radiation
exposure limit on each of the P/T curves for
SSES Units 1 and 2.
The P/T limits are prescribed during all
operational conditions to avoid encountering
pressure, temperature, and temperature rate
of change conditions that might cause
undetected flaws to propagate, resulting in
nonductile failure of the reactor coolant
pressure boundary, an unanalyzed condition.
Therefore, the proposed changes do not have
any effect on the probability of an accident
previously evaluated.
The P/T curves are used as operational
limits during heatup or cooldown
maneuvering, when pressure and
temperature indications are monitored and
compared to the applicable curve to
determine that operation is within the
allowable region. The P/T curves provide
assurance that station operation is consistent
with previously evaluated accidents. Thus,
the radiological consequences of an accident
previously evaluated are not increased.
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?
No. The proposed changes do not change
the response of any plant equipment to
transient conditions. The proposed changes
do not introduce any new equipment, modes
of system operation, or failure mechanisms.
Therefore, there are no new types of
failures or new or different kinds of accidents
or transients that could be created by these
changes. The proposed changes do not create
the possibility of a new or different kind of
accident from any accident previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
No. The consequences of a previously
evaluated accident are not increased by these
proposed changes, since the Loss of Coolant
Accident analyzed in the FSAR [Final Safety
Analysis Report] assumes a complete break of
the reactor coolant pressure boundary. The
changes to the P/T limits curves do not
change this assumption.
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
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: Bryan A. Snapp,
Esquire, Assoc. General Counsel, PPL
Services Corporation, 2 North Ninth St.,
GENTW3, Allentown, PA 18101–1179.
NRC Branch Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos.
50–387 and 50–388, Susquehanna
Steam Electric Station, Units 1 and 2
(SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request:
November 9, 2004, as supplemented
December 15, 2005. This notice
supersedes the original notice published
on April 26, 2005 (70 FR 21463), which
was based upon the licensee’s
application dated November 9, 2004.
Description of amendment request:
The proposed amendments would
change the SSES 1 and 2 Technical
Specifications (TSs) 3.8.4, ‘‘DC
Sources— Operating,’’ 3.8.5, ‘‘DC
Sources—Shutdown,’’ 3.8.6, ‘‘Battery
Cell Parameters,’’ and add a new TS
section, 5.5.13, ‘‘Battery Monitoring and
Maintenance Program.’’ These changes
are consistent with Technical
Specification Change Traveler (TSTF)
360, Revision 1.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
No. The proposed changes restructure the
Technical Specifications (TSs) for the DC
Electrical Power Systems. The proposed
changes consist of the relocation of several
surveillance requirements that perform
preventive maintenance on the safety related
batteries, to a new license controlled
program. The DC electrical power systems,
including associated battery chargers, are not
initiators to any accident sequence analyzed
in the Final Safety Analysis Report (FSAR).
Operation in accordance with the proposed
TS ensures that the DC electrical power
systems are capable of performing functions
as described in the FSAR. Therefore, the
mitigative functions supported by the DC
Power Systems will continue to provide the
protection assumed by the analysis.
PO 00000
Frm 00085
Fmt 4703
Sfmt 4703
The relocation of preventive maintenance
surveillance, and certain operating limits and
actions to a newly created, licenseecontrolled TS 5.5.13, ‘‘Battery Monitoring
and Maintenance Program,’’ will not
challenge the ability of the DC electrical
power systems to perform their design
functions. The maintenance and monitoring
required by current TS, which are based on
industry standards, will continue to be
performed. In addition, the DC Power
Systems are 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 systems. The integrity of
fission product barriers, plant configuration,
and operating procedures as described in the
FSAR will not be affected by the proposed
changes.
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?
No. The proposed changes involve
restructuring the TS for the DC electrical
power systems. These changes will rely on a
new license controlled program to monitor
battery parameters for operability. The DC
electrical power systems, which include the
associated battery chargers, are not initiators
to any accident sequence analyzed in the
FSAR. Rather, the DC electrical power
systems are used to supply equipment used
to mitigate an accident. These mitigative
functions, supported by the DC electrical
power systems are not affected by these
changes and they will continue to provide
the protection assumed by the safety analysis
described in the FSAR. There are no new
types of failures or new or different kinds of
accidents or transients that could be created
by these changes.
Therefore, the proposed changes do not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
No. The margin of safety is established
through equipment design, operating
parameters, and the setpoints at which
automatic actions are initiated. The proposed
changes will not adversely affect operation of
plant equipment. These changes will not
result in a change to the setpoints at which
protective actions are initiated. Sufficient DC
electrical system capacity is ensured to
support operation of mitigation equipment.
The changes associated with the new Battery
Maintenance and Monitoring Program will
ensure that the station batteries are
maintained in a highly reliable state. The
equipment fed by the DC electrical sources
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 a margin of
safety.
E:\FR\FM\17JAN1.SGM
17JAN1
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
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: Bryan A. Snapp,
Esquire, Assoc. General Counsel, PPL
Services Corporation, 2 North Ninth St.,
GENTW3, Allentown, PA 18101–1179.
sroberts on PROD1PC69 with NOTICES
Union Electric Company, Docket No.
50–483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October
26, 2005.
Description of amendment request:
The amendment would revise Technical
Specification (TS) 3.6.6, ‘‘Containment
Spray and Cooling Systems,’’ to change
Required Action D.1 that currently
allows 72 hours of operation with both
containment cooling trains out of
service as long as both containment
spray trains are operable. The required
action would be revised to impose the
more stringent requirement of requiring
plant shutdown if both containment
cooling trains are out of service instead
of allowing the 72 hours to restore an
inoperable train. There are also changes
to other required actions in TS 3.6.6 to
reflect the revision to Required Action
D.1. In addition, the required action for
two inoperable containment spray trains
is being revised.
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 in the required
action when two containment cooling trains
are inoperable to require plant shutdown is
more restrictive than the current required
action that allows 72 hours of operation [to
restore one containment cooling train to
operable status]. Also the proposed change to
the required action [F.1 for] when two
containment cooling trains are inoperable to
be in MODE 3 within 6 hours and MODE 5
within 36 hours [are the same as in the
current Required Actions E.1 and E.2 for
when the two containment cooling trains are
inoperable. The proposed change to the
required action for two containment spray
trains being inoperable] is more restrictive
than the current required action to enter LCO
[Limiting Condition for Operation] 3.0.3
immediately [because] LCO 3.0.3 requires the
plant to be in MODE 3 within 7 hours. The
more stringent requirements are imposed to
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
ensure process variables, structures, systems
and components are maintained consistently
with the safety analysis and licensing basis
[for Callaway].
All of these proposed changes have been
reviewed to ensure no previously evaluated
accident has been adversely affected. [The
proposed changes are not accident initiators.]
Therefore, this 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
The proposed change does not involve a
physical alteration of the plant (no new or
different type of equipment will be installed)
or changes in controlling [plant] parameters.
The proposed change does impose different
requirements. However, these changes are
consistent with [the] assumptions made in
the safety analysis and licensing basis [for
Callaway]. Thus, this 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 a margin of safety?
Response: No
The imposition of more stringent
requirements has no impact on or will
increase the margin of safety. The change in
the required action when two containment
cooling trains are out of service will increase
the margin of safety by decreasing the
allowed restoration time [to restore an
inoperable containment cooling train to
operable status].
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: John O’Neill,
Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Branch Chief: David Terao.
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
PO 00000
Frm 00086
Fmt 4703
Sfmt 4703
2597
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 (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
Systems (ADAMS) Public Electronic
Reading Room on the Internet at the
NRC Web site, https://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 PDR
Reference staff at 1 (800) 397–4209,
(301) 415–4737 or by e-mail to
pdr@nrc.gov.
AmerGen Energy Company, LLC, et al.,
Docket No. 50–219, Oyster Creek
Nuclear Generating Station (OCNGS),
Ocean County, New Jersey
Date of application for amendment:
December 17, 2004.
Brief description of amendment: The
amendment revised Appendix B,
Environmental Technical
Specifications, of the OCNGS Facility
Operating License, principally by
deleting redundant reporting
requirements, aligning various
requirements with regulations and
accepted guidance documents, and
correcting administrative errors.
Date of Issuance: January 4, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days of issuance.
E:\FR\FM\17JAN1.SGM
17JAN1
2598
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
Amendment No.: 257.
Facility Operating License No. DPR–
16: The amendment revised the
Environmental Technical
Specifications.
Date of initial notice in Federal
Register: April 12, 2005 (70 FR 19113).
The Commission’s related evaluation
of this amendment is contained in a
Safety Evaluation dated January 4, 2006.
No significant hazards consideration
comments received: No.
Dominion Nuclear Connecticut, Inc.,
Docket No. 50–336, Millstone Power
Station, Unit No. 2, New London
County, Connecticut
Date of application for amendment:
February 25, 2005, as supplemented by
letter dated August 4, 2005.
Brief description of amendment: The
amendment revised the Millstone Power
Station, Unit No. 2, Technical
Specifications Surveillance
Requirement for trisodium phosphate to
remove the granularity term and
chemical detail. In addition, the
proposed change will increase the
allowed outage time from 48 to 72
hours.
Date of issuance: January 3, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 60 days from the date of
issuance.
Amendment No.: 290.
Facility Operating License No. DPR–
65: The amendment revised the
Technical Specifications.
Date of initial notice in Federal
Register: July 19, 2005 (70 FR 41444).
The additional information provided in
the supplemental letter dated August 4,
2005, did not expand the scope of the
application as noticed and did not
change the NRC staff’s original proposed
no significant hazards consideration
determination.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated January 3, 2006.
No significant hazards consideration
comments received: No.
sroberts on PROD1PC69 with NOTICES
Dominion Nuclear Connecticut, Inc., et
al., Docket No. 50–423, Millstone Power
Station, Unit No. 3, New London
County, Connecticut
Date of application for amendment:
December 16, 2004, as supplemented on
October 5, 2005.
Brief description of amendment: The
amendment revised the current fuel rod
average licensing basis burnup limit for
one lead test assembly containing
advanced zirconium based alloys to a
limit not exceeding 71,000 megawattdays per metric ton of uranium.
Date of issuance: December 30, 2005.
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
Effective date: As of the date of
issuance and shall be implemented
within 60 days from the date of
issuance.
Amendment No.: 228
Facility Operating License No. NPF–
49: The amendment revised the design
basis.
Date of initial notice in Federal
Register: February 1, 2005 (70 FR
5238). The October 5, 2005, supplement
provided clarifying information and did
not change the initial proposed no
significant hazards consideration
determination.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated December 30,
2005.
No significant hazards consideration
comments received: No.
Entergy Operations, Inc., Docket No.
50–382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish,
Louisiana
Date of amendment request: July 20,
2005, as supplemented by letter dated
September 14, 2005.
Brief description of amendment: The
amendment approves the transfer of
Facility Operating License and Materials
License No. NPF–38, held by Entergy
Louisiana, Inc. (ELI) and Entergy
Operatings, Inc. (EOI), for the Waterford
Steam Electric Station, Unit 3
(Waterford 3). The transfer is associated
with the restructuring of ELI from a
Louisiana corporation to a Texas limited
liability company, Entergy Louisiana,
LLC (ELL). EOI will continue to operate
Waterford 3, and the restructuring will
not affect the technical or financial
qualifications of ELL or EOI.
Date of issuance: December 31, 2005.
Effective date: At the time the transfer
is completed.
Amendment No.: 203.
Facility Operating License No. NPF–
38: The amendment revised the Facility
Operating License and Materials
License.
Date of initial notice in Federal
Register: October 17, 2005 (70 FR
60374).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated December 2,
2005.
No significant hazards consideration
comments received: No.
Exelon Generation Company, LLC, and
PSEG Nuclear LLC, Docket Nos. 50–277
and 50–278, Peach Bottom Atomic
Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of application for amendments:
December 17, 2004.
PO 00000
Frm 00087
Fmt 4703
Sfmt 4703
Brief description of amendments: The
amendments revised the Appendix B,
Environmental Technical
Specifications.
Date of issuance: January 3, 2006.
Effective date: As of the date of
issuance, to be implemented within 60
days.
Amendments Nos.: 257 and 260.
Renewed Facility Operating License
Nos. DPR–44 and DPR–56: The
amendments revised the Environmental
Technical Specifications.
Date of initial notice in Federal
Register: April 12, 2005 (70 FR 19112).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated January 3, 2006.
No significant hazards consideration
comments received: No.
FPL Energy Seabrook, LLC, Docket No.
50–443, Seabrook Station, Unit No. 1,
Rockingham County, New Hampshire
Date of amendment request: March
28, 2005, as supplemented September
23, 2005.
Description of amendment request:
The amendment extended the
expiration of Facility Operating License
(FOL) NPF–86 for Seabrook Station,
Unit No. 1, by approximately 3.4 years.
The extension sets the date of expiration
of the FOL to occur 40 years from the
date of issuance of the full-power
operating license. Specifically, the FOL,
with a previous expiration date of
October 17, 2026, now expires March
15, 2030. This change allows the
recapture of zero-power and low-power
testing time in accordance with SECY–
98–296, ‘‘Agency Policy Regarding
Licensee Recapture of Low-Power
Testing or Shutdown Time for Nuclear
Power Plants,’’ dated December 21,
1998.
Date of issuance: December 28, 2005.
Effective date: As of its date of
issuance, and shall be implemented
within 30 days.
Amendment No.: 105.
Facility Operating License No. NPF–
86: The amendment revised the License.
Date of initial notice in Federal
Register: May 24, 2005 (70 FR 29797).
The licensee’s September 23, 2005
supplement provided clarifying
information that did not change the
scope of the proposed amendment as
described in the original notice of
proposed action published in the
Federal Register, and did not change
the initial proposed no significant
hazards consideration determination.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated December 28,
2005.
No significant hazards consideration
comments received: No.
E:\FR\FM\17JAN1.SGM
17JAN1
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
Florida Power and Light Company, et
al., Docket No. 50–389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of application for amendment:
March 31, 2005, as supplemented
November 9, 2005.
Brief description of amendment: This
amendment extended the date for the
next Appendix J, Type A test at St.
Lucie Unit 2 until the end of the SL2–
17 refueling outage.
Date of Issuance: December 23, 2005.
Effective Date: As of the date of
issuance and shall be implemented
within 60 days.
Amendment No.: 140.
Renewed Facility Operating License
No. NPF–16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal
Register: June 7, 2005 (70 FR 33215).
The November 9, 2005, supplement did
not affect the original proposed no
significant hazards determination, or
expand the scope of the request as
noticed in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated December 23,
2005.
No significant hazards consideration
comments received: No.
sroberts on PROD1PC69 with NOTICES
Pacific Gas and Electric Company,
Docket Nos. 50–275 and 50–323, Diablo
Canyon Nuclear Power Plant, Unit Nos.
1 and 2, San Luis Obispo County,
California
Date of application for amendments:
October 29, 2004, as supplemented by
letters dated May 6 and October 31,
2005.
Brief description of amendments: The
amendments revised the Technical
Specification (TS) requirements for the
handling of irradiated fuel in the
containment and fuel building, and
certain specifications related to
performing core alterations. These
changes are based on analysis of the
postulated fuel handling and core
alteration accidents and transients for
Diablo Canyon Nuclear Power Plant,
Units 1 and 2. The amendments are
consistent with the NRC-approved
Industry/Technical Specification Task
Force (TSTF) Standard Technical
Specifications Change Traveler, TSTF–
51, Revision 2, ‘‘Revise containment
requirements during handling irradiated
fuel and core alterations.’’ In addition,
the amendments made editorial
corrections to TS 3.1.7, ‘‘Rod Position
Indication,’’ TS 3.3.1, ‘‘Reactor Trip
System (RTS) Instrumentation,’’ TS
3.4.16, ‘‘RCS Specific Activity,’’ TS
3.7.3, ‘‘Main Feedwater Isolation Valve
(MFIVs), Main Feedwater Regulating
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
Valves (MFRVs), MFRV Bypass Valves,
and Main Feedwater Pump (MFWP)
Turbine Stop Valves,’’ and TS 3.7.13,
‘‘Fuel Handling Building Ventilation
System (FHBVS).’’
Date of issuance: January 3, 2006.
Effective date: January 3, 2006, and
shall be implemented within 90 days of
issuance.
Amendment Nos.: Unit 1—184; Unit
2—86.
Facility Operating License Nos. DPR–
80 and DPR–82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal
Register: January 4, 2005 (70 FR 403)
The supplements dated May 6 and
October 31, 2005, 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
no significant hazards consideration
determination.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated January 3, 2006.
No significant hazards consideration
comments received: No.
Southern Nuclear Operating Company,
Inc., Georgia Power Company,
Oglethorpe Power Corporation,
Municipal Electric Authority of
Georgia, City of Dalton, Georgia, Docket
Nos. 50–321 and 50–366, Edwin I.
Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments:
November 12, 2004, as supplemented by
letters dated September 2 and
September 16, 2005.
Brief description of amendments: The
amendments revised the Technical
Specifications (TS) 3.1.7, ‘‘Standby
Liquid Control (SLC) System,’’ for
Hatch, Units 1 and 2. The amendments
update Figure 3.1.7–1 and 3.1.7–2 of the
Units 1 and 2 TS to reflect the increased
concentration of Boron-10 in the
solution. Conforming revisions to Bases
B3.1.7, are also included.
Date of issuance: January 5, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 30 days from the date of
issuance.
Amendment Nos.: 247/191.
Renewed Facility Operating License
Nos. DPR–57 and NPF–5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal
Register: February 1, 2005 (70 FR
5249).
The supplemental letter dated
September 2, 2005, contained clarifying
information only and did not change the
initial proposed no significant hazards
consideration determination or expand
PO 00000
Frm 00088
Fmt 4703
Sfmt 4703
2599
the scope of the original Federal
Register notice. The supplemental letter
dated September 16, 2005, contained
information that expanded the scope of
the original Federal Register notice. The
proposed amendment was re-noticed on
October 25, 2005 (70 FR 61662).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated January 5, 2006.
No significant hazards consideration
comments received: No.
Tennessee Valley Authority, Docket
Nos. 50–327 and 50–328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton
County, Tennessee
Date of application for amendments:
April 27, 2005, as supplemented by
letter dated November 17, 2005.
Brief description of amendments: The
amendments relocate several Technical
Specification (TS) requirements to the
Sequoyah Technical Requirements
Manual (TRM). Specifically, the
amendments relocate the provisions for
TS 3.3.2 (Movable Incore Detectors), TS
3.3.3.4 (Meteorological
Instrumentation), TS 3.4.7 (Reactor
Coolant System Chemistry), TS 3.4.11
(Reactor Coolant System Head Vents),
TS 3.7.2 (Steam Generator Pressure and
Temperature Limitations), TS 3.7.10
(Sealed Source Contamination), TS 3.9.5
(Refueling Operations
Communications), and TS 3.9.6
(Manipulator Crane) to the TRM. These
changes are consistent with the latest
version of NUREG–1431, Revision 3,
‘‘Standard Technical Specifications for
Westinghouse Plants,’’ and do not
diminish the level of safety found in the
current TSs.
Date of issuance: December 28, 2005.
Effective date: As of the date of
issuance and shall be implemented
within 45 days.
Amendment Nos.: 305, 295.
Facility Operating License Nos. DPR–
77 and DPR–79: Amendments revised
the technical specifications.
Date of initial notice in Federal
Register: July 5, 2005 (70 FR 38723).
The supplemental letter of November
17, 2005, provided clarifying
information that did not change the
initial proposed no significant hazards
consideration determination.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated December 28,
2005.
No significant hazards consideration
comments received: No.
E:\FR\FM\17JAN1.SGM
17JAN1
2600
Federal Register / Vol. 71, No. 10 / Tuesday, January 17, 2006 / Notices
Virginia Electric and Power Company,
et al., Docket Nos. 50–280 and 50–281,
Surry Power Station, Unit Nos. 1 and 2,
Surry County, Virginia
Date of application for amendments:
December 17, 2004.
Brief Description of amendments:
These amendments revised the reactor
coolant pressure and temperature limits,
low-temperature overpressure
protection system (LTOPS) setpoint
values, and LTOPS enable temperatures
that are valid for up to 47.6 effective
full-power years (EFPY) and 48.1 EFPY
of operation at Surry Power Station,
Unit Nos. 1 and 2, respectively.
Date of issuance: January 3, 2006.
Effective date: As of the date of
issuance and shall be implemented
within 180 days from the date of
issuance.
Amendment Nos.: 245/244.
Renewed Facility Operating License
Nos. DPR–32 and DPR–37: Amendments
change the Technical Specifications.
Date of initial notice in Federal
Register: March 1, 2005 (70 FR 9999).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated January 3, 2006.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 9th day
of January 2006.
For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Deputy Director, Division of Operating
Reactor Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 06–320 Filed 1–13–06; 8:45 am]
BILLING CODE 7590–01–P
Written comments regarding
OMB’s Proposed Risk Assessment
Bulletin are due by June 15, 2006. This
date has been selected in order to permit
the public to participate in a related
workshop to be organized by the NAS,
prior to submitting their written
comments.
DATES:
Because of potential delays
in OMB’s receipt and processing of
mail, respondents are strongly
encouraged to submit comments
electronically to ensure timely receipt.
We cannot guarantee that comments
mailed will be received before the
comment closing date. Electronic
comments may be submitted to:
OMB_RAbulletin@omb.eop.gov. Please
put the full body of your comments in
the text of the electronic message and as
an attachment. Please include your
name, title, organization, postal address,
telephone number and e-mail address in
the text of the message. Please be aware
that all comments are available for
public inspection. Accordingly, please
do not submit comments containing
trade secrets, confidential or proprietary
commercial or financial information, or
other information that you do not want
to be made available to the public.
Comments also may be submitted via
facsimile to (202) 395–7245.
FOR FURTHER INFORMATION CONTACT: Dr.
Nancy Beck, Office of Information and
Regulatory Affairs, Office of
Management and Budget, 725 17th
Street, NW., New Executive Office
Building, Room 10201, Washington, DC
20503. Telephone (202) 395–3093.
ADDRESSES:
OMB is
seeking comments on its Proposed Risk
Assessment Bulletin by June 15, 2006.
The proposed Risk Assessment Bulletin
is posted on OMB’s Web site, https://
www.whitehouse.gov/omb/inforeg/
infopoltech.html#iq.
SUPPLEMENTARY INFORMATION:
OFFICE OF MANAGEMENT AND
BUDGET
Proposed Risk Assessment Bulletin
Office of Management and
Budget.
ACTION: Notice of proposed Bulletin and
request for comments.
sroberts on PROD1PC69 with NOTICES
AGENCY:
SUMMARY: As part of an ongoing effort to
improve the quality, objectivity, utility,
and integrity of information
disseminated by the Federal
Government to the public, the Office of
Management and Budget (OMB), in
consultation with the Office of Science
and Technology Policy (OSTP), has
referred to the National Academy of
Sciences (NAS), for their expert review,
new guidance to enhance the quality
and objectivity of risk assessments
produced by the Federal Government.
OMB will also be accepting public
comment on this document until June
15, 2006.
VerDate Aug<31>2005
15:57 Jan 13, 2006
Jkt 208001
John D. Graham,
Administrator, Office of Information and
Regulatory Affairs.
[FR Doc. E6–345 Filed 1–13–06; 8:45 am]
BILLING CODE 3110–01–P
PENSION BENEFIT GUARANTY
CORPORATION
Exemption From the Bond/Escrow
Requirement Relating to the Sale of
Assets by an Employer Who
Contributes to a Multiemployer Plan;
LA Team Co. LLC
Pension Benefit Guaranty
Corporation.
ACTION: Notice of exemption.
AGENCY:
PO 00000
Frm 00089
Fmt 4703
Sfmt 4703
SUMMARY: The Pension Benefit Guaranty
Corporation has granted a request from
the LA Team Co. LLC for an exemption
from the bond/escrow requirement of
section 4204(a)(1)(B) of the Employee
Retirement Income Security Act of 1974,
as amended, with respect to the Major
League Baseball Players Pension Plan. A
notice of the request for exemption from
the requirement was published on July
7, 2005 (70 FR 39349). The effect of this
notice is to advise the public of the
decision on the exemption request.
ADDRESSES: The non-confidential
portions of the request for an exemption
and the PBGC response to the request
may be obtained by writing PBGC’s
Communications and Public Affairs
Department (‘‘CPAD’’) at Suite 1200,
1200 K Street, NW., Washington, DC
20005–4026, or by visiting or calling
CPAD (202–326–4040) during normal
business hours.
FOR FURTHER INFORMATION CONTACT:
Gennice D. Brickhouse, Office of the
Chief Counsel, Suite 340, 1200 K Street,
NW., Washington, DC 20005–4026;
telephone 202–326–4020. (For TTY/
TDD users, call the Federal Relay
Service toll-free at 1–800–877–8339 and
ask to be connected to 202–326–4020).
SUPPLEMENTARY INFORMATION:
Background
Section 4204 of the Employee
Retirement Income Security Act of 1974,
as amended by the Multiemployer
Pension Plan Amendments Act of 1980
(‘‘ERISA’’ or ‘‘the Act’’), provides that a
bona fide arm’s-length sale of assets of
a contributing employer to an unrelated
party will not be considered a
withdrawal if three conditions are met.
These conditions, enumerated in section
4204(a)(1)(A)–(C), are that:
(A) The purchaser has an obligation to
contribute to the plan with respect to
the operations for substantially the same
number of contribution base units for
which the seller was obligated to
contribute;
(B) The purchaser obtains a bond or
places an amount in escrow, for a period
of five plan years after the sale, in an
amount equal to the greater of the
seller’s average required annual
contribution to the plan for the three
plan years preceding the year in which
the sale occurred or the seller’s required
annual contribution for the plan year
preceding the year in which the sale
occurred (the amount of the bond or
escrow is doubled if the plan is in
reorganization in the year in which the
sale occurred); and
(C) The contract of sale provides that
if the purchaser withdraws from the
plan within the first five plan years
E:\FR\FM\17JAN1.SGM
17JAN1
Agencies
[Federal Register Volume 71, Number 10 (Tuesday, January 17, 2006)]
[Notices]
[Pages 2586-2600]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-320]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and 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 December 22, 2005 to January 5, 2006. The
last biweekly notice was published on January 3, 2006 (71 FR 145).
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. Within 60 days after the date of publication of this
notice, 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
[[Page 2587]]
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.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place 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
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, 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 2.309, 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov;
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. 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 it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-
[[Page 2588]]
mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and
petition for leave to intervene should also be sent to the attorney for
the licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
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 ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, https://
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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to pdr@nrc.gov.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: December 2, 2005.
Description of amendment request: The amendment would revise the
Technical Specifications to increase the allowable as-found main steam
safety valve code safety function lift setpoint tolerance from 1% to 3%.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes allow for an increase in the as-found
Main Steam Safety Valve (MSSV) setpoint tolerance from 1% to 3%. The proposed changes do not alter the
MSSV nominal lift setpoints or MSSV lift setpoint test frequency.
The proposed TS changes have been evaluated on both a generic
and plant specific basis. The NRC has approved the general approach
of this change; however, implementation is contingent on several
plant specific evaluations. The required plant specific analyses and
evaluations included transient analysis of the anticipated
operational transients (AOTs); analysis of the design basis
overpressurization event; evaluation of the performance of high
pressure systems, and evaluation of the containment response during
Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the MSSV
discharge lines and containment. These analyses and evaluations
demonstrate that there is adequate margin to the design core thermal
limits and reactor vessel pressure limits using the 3%
MSSV as-found setpoint tolerance. The analyses and evaluations also
demonstrate that the operation of high-pressure safety systems will
not be adversely affected and that the containment response during a
LOCA will be acceptable.
Evaluations of the impact of the proposed change on the
equipment important to safety have been performed and no adverse
conditions were identified. The reactor pressure vessel and attached
systems and piping have been evaluated for the impact of this
proposed TS change. A plant specific analysis has been performed
which indicates that the ASME Code upset limits for the reactor
pressure vessel will not be exceeded for the limiting event, i.e.,
Main Steam Isolation Valve (MSIV) closure with flux Scram. The
reactor pressure vessel and attached piping design values will not
be exceeded. Therefore, the probability of a malfunction of the
reactor pressure vessel and attached systems and piping is not
increased and the consequences of such an accident remain
acceptable.
The nuclear fuel has been evaluated for the impact of the
proposed change.
Plant specific analyses were performed which indicate that for
all abnormal operational transients adequate margin to the fuel
thermal limit parameters, i.e., Minimum Critical Power Ratio (MCPR)
and thermal-mechanical limits, is maintained. Emergency Core Cooling
System (ECCS)/LOCA performance is maintained adequate to meet the
requirements of 10 CFR 50.46. Therefore, the consequences of these
accidents remain acceptable and the probability of the malfunction
of the nuclear fuel is not increased.
The Containment response during a LOCA has been evaluated for
the impact of the proposed change. The major factor in the
Containment pressure response to a LOCA is the rate of reactor
vessel water inventory loss due to a DBA LOCA. The rate of reactor
vessel water inventory loss is mainly dependent on the initial
reactor pressure, which is not affected by the proposed setpoint
tolerance change. The major factor in the Containment temperature
response to a LOCA is the integrated steam inventory loss due to
Main Steamline Break. The rate of reactor vessel steam inventory
loss is mainly dependent on the reactor decay heat, which is not
affected by the proposed setpoint tolerance change. Therefore, the
consequences of these accidents remain acceptable and the
probability of the malfunction of Containment is not increased.
The Control Rod Drive (CRD) system has been evaluated for the
impact of the proposed change. The CRD system capability of
controlling reactor power during normal plant operation and rapidly
inserting control rod blades (Scram) during abnormal plant
conditions is not impacted by the proposed change. Therefore, the
probability of a malfunction of the CRD system is not increased.
The Reactor Vessel Instrumentation System has been evaluated for
the impact of the proposed change. The Reactor Vessel
Instrumentation System will continue to be operated within the
current design pressure/temperature requirements; therefore, the
probability of a malfunction of the Reactor Vessel Instrumentation
System is not increased.
An administrative change is also being proposed to correct the
reference to ``IWV-3510 of Section XI of the ASME Boiler and
Pressure Vessel Code'' in TS 4.3.E because the stated ASME section
no longer exists. The TS is being changed to reference specification
4.3.C for MSSV testing. This is an administrative change and does
not affect previously evaluated accidents.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. Will operation of the facility in accordance of the proposed
amendment create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed TS changes allow for an increase in the as-found
MSSV setpoint tolerance from 1% to 3%.
Generic and plant specific analyses and evaluations indicate that
the plant response to any previously evaluated event will remain
acceptable. All plant systems, structures, and components will
continue to be capable of performing their required safety function
as required by event analysis guidance.
The proposed TS changes do not alter the MSSV nominal lift
setpoints or MSSV lift setpoint test frequency. The operation and
response of the affected equipment important to safety is unchanged.
All systems, structures, and components will continue to be operated
within acceptable operating and/or design parameters. No system,
structure, or component will be subjected to a condition that has
not been evaluated and determined to be acceptable using the
guidance required for specific event analysis.
The change to correct the reference to ``IWV-3510 of Section XI
of the ASME Boiler and Pressure Vessel Code'' in TS 4.3.E is an
administrative change and does not affect the possibility of a new
or different kind of accident.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
identified.
3. Will operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The proposed TS changes allow for an increase in the as-found
MSSV setpoint tolerance from 1% to 3%. The
proposed TS changes do not alter the MSSV nominal lift setpoints or
MSSV lift setpoint test
[[Page 2589]]
frequency. The operation and response of the affected equipment
important to safety is unchanged. All systems, structures, and
components will continue to be operated within acceptable operating
and/or design parameters. While the calculated peak reactor vessel
pressure for the ASME overpressure event is higher than that
calculated without the increase in setpoint tolerance, it is still
within the respective licensing acceptance limits associated with
this event. These licensing acceptance limits have been determined
by the NRC to provide a sufficient margin of safety.
The increase in MSSV steam flow and reactor vessel pressure does
not reduce the margin of safety associated with the MSSVs and
associated components and structures since the increased MSSV steam
flow rate and reactor vessel pressure are bounded by the current
design analysis.
The margin of safety for fuel thermal limits and 10 CFR 50.46
limits are unaffected by the proposed change.
The margin of safety for the Containment is unaffected by the
proposed change.
The capability of the SLC system and the CRD system to perform
their safety functions during all required events, using the
required guidance for event analysis, is maintained. Therefore, the
proposed changes do not reduce the margin of safety provided by the
SLC and CRD systems.
The change to correct the reference to ``IWV-3510 of Section XI
of the ASME Boiler and Pressure Vessel Code'' in TS 4.3.E is an
administrative change and does not affect the margin of safety.
Therefore, these proposed TS 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: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Darrell J. Roberts.
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: November 3, 2005.
Description of amendments request: The proposed amendments would
revise the accident source term in the design-basis radiological
consequences analyses and the associated Technical Specifications
(TSs), pursuant to section 50.67 of part 50 of Title 10 of the Code of
Federal Regulations (10 CFR 50.67). The proposed amendments would
provide for the full implementation of the alternate source term (AST)
in accordance with the guidance in Regulatory Guide 1.183,
``Alternative Radiological Source Terms for Evaluating Design Basis
Accidents at Nuclear Power Reactors.'' The proposed amendments would
also increase the flow rate for the control room emergency ventilation
system (CREVS) from 2000 to 10000 cubic feet per minute in TS 5.5.11,
``Ventilation Filter Testing Program,'' by means of a modification to
the CREVS. In addition, automatic isolation dampers and radiation
monitors will also be installed at access control heating, ventilating,
and air conditioning (HVAC) unit no. RTU-1 and access control air
conditioning unit no. 13.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The results of the applicable radiological design basis
accidents (DBAs) re-evaluation demonstrated that, with the requested
changes, the dose consequences of these limiting events are within
the regulatory limits and guidance provided by the Nuclear
Regulatory Commission in 10 CFR 50.67 and Regulatory Guide 1.183 for
AST methodology. The AST is an input to calculations used to
evaluate the consequences of an accident and does not by itself
affect the plant response or the actual pathway of the activity
released from the fuel. It does, however, better represent the
physical characteristics of the release such that appropriate
mitigation techniques may be applied.
The change from the original source term to the new proposed AST
is a change in the analysis method and assumptions and has no effect
on accident initiators or causal factors that contribute to the
probability of occurrence of previously analyzed accidents. Use of
an AST to analyze the dose effect of DBAs shows that regulatory
acceptance criteria for the new methodology continues to be met.
Changing the analysis methodology does not change the sequence or
progression of the accident scenario.
The proposed Technical Specification changes reflect the plant
configuration that will either support implementation of the AST
analyses or eliminate requirements that are no longer needed as a
result of the revised DBA analyses. The equipment affected by the
proposed changes is mitigative in nature and relied upon after an
accident has been initiated. The operation of various filtration
systems have been considered in the evaluations for these proposed
changes. While the operation of some systems does change with the
implementation of an AST, the affected systems are not accident
initiators; and application of the AST methodology, itself, is not
an initiator of a DBA.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
As described in Item 1 above, the changes proposed in this
license amendment request involve the use of a new analysis
methodology and related regulatory acceptance criteria. The proposed
Technical Specification changes reflect the plant configuration that
will either support implementation of the new methodology or
eliminate requirements that are no longer needed as a result of the
new methodology. No new or different accidents result from utilizing
the proposed changes. Although the proposed changes require
modification to the Control Room emergency ventilation system and
installation of automatic isolation dampers and radiation monitors
at Access Control HVAC Unit RTU-1 and Access Control Air
Conditioning Unit 13 on the Auxiliary Building roof, none of these
changes can initiate a new or different kind of accident since they
are only related to system capabilities that provide protection from
accidents that have already occurred. As a result, no new failure
modes are being introduced that could lead to different accidents.
These changes do not alter the nature of events postulated in the
Updated Final Safety Analysis Report nor do they introduce any
unique precursor mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
As described in Item 1 above, the changes proposed in this
license amendment request involve the use of a new analysis
methodology and related regulatory acceptance criteria. The proposed
Technical Specification changes reflect the plant configuration that
will either support implementation of the new methodology or
eliminate requirements that are no longer needed as a result of the
new methodology. Safety margins and analytical conservatisms have
been evaluated and have been found acceptable. The analyzed events
have been carefully selected and, with plant modification, margin
has been retained to ensure that the analyses adequately bound
postulated event scenarios. The analyses have been performed using
conservative methodologies, as specified in Regulatory Guide 1.183.
The dose consequences of these DBAs remain within the acceptance
criteria presented in 10 CFR 50.67, ``Accident Source Term,'' and
Regulatory Guide 1.183. The proposed changes continue to ensure that
the doses at the exclusion area boundary and low population zone
boundary, as well as the Control Room, are within corresponding
regulatory limits.
[[Page 2590]]
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 amendments request involves no significant hazards
consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt
Street, 17th floor, Baltimore, MD 21202.
NRC Branch Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of amendment request: December 14, 2005.
Description of amendment request: The proposed amendment
modifies the Technical Specifications (TSs) to incorporate a revised
Single Loop Operation Safety Limit Minimum Critical Power Ratio (SLO
SLMCPR) due to the cycle-specific analysis.
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 derivation of the cycle specific Single Loop Operation
Safety Limit Minimum Critical Power Ratio (SLO SLMCPR) for
incorporation into the Technical Specifications (TS), and its use to
determine cycle-specific thermal limits, has been performed using
the methodology discussed in ``General Electric Standard Application
for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and U.S.
Supplement, NEDE-24011-P-A-15-US, September, 2005, which includes
Amendment 25. Amendment 25 was approved by the NRC in a March 11,
1999 safety evaluation report.
The basis of the SLO SLMCPR calculation is to ensure that
greater than 99.9% of all fuel rods in the core avoid transition
boiling if the limit is not violated. The new SLO SLMCPR preserves
the existing margin to transition boiling. The GE-14 fuel is in
compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-15-US, September 2005, which
provides the fuel licensing acceptance criteria. The probability of
fuel damage will not be increased as a result of this change.
Therefore, the proposed TS 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.
The SLO SLMCPR is a TS numerical value, calculated to ensure
that transition boiling does not occur in 99.9% of all fuel rods in
the core if the limit is not violated. The new SLO SLMCPR is
calculated using NRC approved methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-15
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September
2005, which includes Amendment 25. Additionally, the GE-14 fuel is
in compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, which
provides the fuel licensing acceptance criteria. The SLO SLMCPR is
not an accident initiator, and its revision will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
There is no significant reduction in the margin of safety
previously approved by the NRC as a result of the proposed change to
the SLO SLMCPR, which includes the use of GE-14 fuel. The new SLO
SLMCPR is calculated using methodology discussed in ``General
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-15
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-15-US, September,
2005, which includes Amendment 25. The SLO SLMCPR ensures that
greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated when all
uncertainties are considered, thereby preserving the fuel cladding
integrity.
Therefore, the proposed TS change will not involve a significant
reduction in [a] margin of safety previously approved by the NRC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Brad Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Darrell J. Roberts.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: December 21, 2005.
Description of amendment request: The proposed amendment revises
the Technical Specifications by relocating the Pressure Isolation
Valve (PIV) tables to the Technical Requirements Manual (TRM).
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 relocation of Technical Specification Table
3.4.3.2-1 does not alter the requirements for pressure isolation
valve operability or surveillance currently in the Technical
Specifications. The proposed change to remove the pressure isolation
valve table from TS and relocate the information to an
administratively controlled document, and to revise the wording in
TS to reflect this change, will have no impact on any safety related
structures, systems or components. The probability of occurrence of
a previously evaluated accident is not increased because this change
does not introduce any new potential accident initiating conditions.
The consequences of accidents previously evaluated in the UFSAR
[Updated Final Safety Analysis Report] are not affected because the
ability of the PIVs to limit leakage through these valves in amounts
that do not compromise safety is not affected. 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 changes are administrative in nature and do not
result in physical alterations or changes in the method by which any
safety related system performs its intended function(s). The
proposed changes do not impact any safety analysis assumptions. The
proposed changes do not create any new accident initiators or
involve an activity that could be an initiator of an accident of a
different type.
All PIVs and alarm instrumentation will continue to be tested to
the same rigorous requirements as defined in the Technical
Specification Surveillance Requirements. The proposed revision does
not make changes in any method of testing or how any safety related
system performs its safety functions. Therefore, the possibility of
an accident of a different type than any previously evaluated in the
UFSAR is not created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The administrative change to relocate Technical Specification
Table 3.4.3.2-1 to the Technical Requirements Manual does not alter
the basic regulatory requirement for Reactor Coolant System pressure
isolation and will not affect the isolation capability for credible
accident scenarios. Future revisions to the Technical Requirements
Manual Table will be subject to evaluation pursuant to 10 CFR 50.59.
Additionally, the proposed relocation does not alter the
requirements for pressure isolation valve and alarm instrumentation
operability currently in the Technical Specifications. The LCO
[limiting condition for operation] and Surveillance Requirements
will be retained in the revised Technical Specifications. The
proposed change will not affect the meaning, application, and
function of the current Technical Specification requirements for the
valves in Table 3.4.3.2-1. Therefore, the proposed changes do not
result in a significant reduction in [a] margin of safety.
[[Page 2591]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Darrell J. Roberts.
Exelon Generation Company, LLC (EGC, licensee), Docket No. 50-265, Quad
Cities Nuclear Power Station (QCNPS), Unit 2, Rock Island County,
Illinois
Date of amendment request: December 15, 2005.
Description of amendment request: The proposed change revises the
values of the safety limit minimum critical power ratio (SLMCPR) in
Technical Specification (TS) section 2.1.1, ``Reactor Core SLs.''
Specifically, the proposed change would require that for Unit 2, the
minimum critical power ratio (MCPR) for Global Nuclear Fuel (GNF) fuel
shall be >=1.09 for two recirculation loop operation, or >=1.10 for
single recirculation loop operation. Additionally, the proposed change
would require that MCPR for Westinghouse fuel shall be >=1.11 for two
recirculation loop operation, or >=1.13 for single recirculation loop
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:
According to 10 CFR 50.92, ``Issuance of amendment,'' paragraph
(c), a proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated; or
(3) Involve a significant reduction in a margin of safety.
EGC has evaluated the proposed change to the TS for QCNPS, Unit
2, using the criteria in 10 CFR 50.92, and has determined that the
proposed change does not involve a significant hazards
consideration. The following information is provided to support a
finding of no significant hazards consideration.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change conservatively establishes the SLMCPR for QCNPS, Unit 2,
Cycle 19 such that the fuel is protected during normal operation and
during plant transients or anticipated operational occurrences
(AOOs).
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during plant transients or AOOs.
Operational limits will be established based on the proposed SLMCPR
to ensure that the SLMCPR is not violated. This will ensure that the
fuel design safety criterion (i.e., that at least 99.9% of the fuel
rods do not experience transition boiling during normal operation
and AOOs) is met. Since the proposed change does not affect
operability of plant systems designed to mitigate any consequences
of accidents, the consequences of an accident previously evaluated
are not expected to increase.
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.
Creation of the possibility of a new or different kind of
accident would require creating one or more new accident precursors.
New accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The proposed change does not involve any plant configuration
modifications or changes to allowable modes of operation. The
proposed change to the SLMCPR assures that safety criteria are
maintained for QCNPS, Unit 2, Cycle 19.
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 SLMCPR provides a margin of safety by ensuring that at least
99.9% of the fuel rods do not experience transition boiling during
normal operation and AOOs if the MCPR limit is not violated. The
proposed change will ensure the appropriate level of fuel protection
by continuing to ensure that at least 99.9% of the fuel rods do not
experience transition boiling during normal operation and AOOs if
the MCPR limit is not violated. Additionally, operational limits
will be established based on the proposed SLMCPR to ensure that the
SLMCPR is not violated. This will ensure that the fuel design safety
criteria (i.e., that no more than 0.1% of the rods are expected to
be in boiling transition if the MCPR limit is not violated) are met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based upon the above, EGC concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of no
significant hazards consideration is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Acting Branch Chief: Mindy S. Landau.
First Energy Nuclear Operating Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio
Date of amendment request: November 21, 2005.
Description of amendment request: The proposed amendment would
revise the acceptance criteria of Technical Specification (TS)
Surveillance Requirements (SRs) associated with TS 3.8.1, ``AC
Sources--Operating,'' to modify the Emergency Diesel Generator (EDG)
start tests to provide minimum voltage and frequency limits and clarify
other limits as steady state parameters. Specifically, the amendment
would revise SRs 3.8.1.2, 3.8.1.7, 3.8.1.12, 3.8.1.15 and 3.8.1.20.
This change is consistent with the approved Technical Specification
Task Force Traveler (TSTF) 163, Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change is a LAR (license amendment request) that
modifies the acceptance criteria for the PNPP TS SRs pertaining to
the EDGs. The EDGs mitigate the consequences of previously evaluated
[[Page 2592]]
accidents involving a loss of offsite power. The EDGs are used to
support mitigation of the consequences of an accident, but they are
not considered as the initiator of any previously analyzed accident.
The proposed LAR does not change the manner in which the EDGs
are operated and when implemented will continue to ensure the EDGs
perform their function when called upon. The proposed revision to
the TS SRs will continue to ensure that minimum frequency and
voltage are attained within the required time. The SRs will continue
to ensure that proper steady state voltage and frequency are
attained consistent with proper EDG governor and voltage regulator
performance.
The proposed LAR does not affect the design of the EDGs, the
operational characteristics of the EDGs, the interfaces between the
EDGs and other plant systems, the function, or reliability of the
EDGs. Thus, the EDGs will be capable of performing their accident
mitigation function and there is no impact to the radiological
consequences of any accident analysis.
As such, the proposed change continues to provide adequate
assurance of operable EDGs and does not involve any increase to the
probability or consequences of an accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed LAR introduces no new mode of plant operation and
it does not involve physical modification to the plant. New
equipment is not installed with the proposed LAR, nor does the
proposed LAR cause existing equipment to be operated in a new or
different manner.
Since the proposed changes do not involve a change to the plant
design or operation, no new system interactions are created by this
change. The proposed LAR does not produce any parameters or
conditions that could contribute to the initiation of accidents
different from those already evaluated in the Updated Safety
Analysis Report.
The changes to the affected TS SRs do not affect the assumed
accident performance of the EDGs, nor any plant structure, system or
component previously evaluated.
Therefore, the proposed LAR does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The proposed change is a LAR that does not impact EDG
performance, including the capability for each EDG to attain and
maintain required voltage and frequency for accepting and supporting
plant safety loads within the required time, as assumed in the plant
safety analysis.
The proposed LAR does not involve a significant reduction in a
margin of safety since the operability of the EDGs continues to be
determined as required to support the capability of the EDGs to
provide emergency power to plant equipment that mitigate the
consequences of an accident.
The proposed LAR does not introduce changes to setpoints or
limits established or assumed by the accident analysis. Therefore,
implementation of the proposed LAR 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Mindy Landau, Acting.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: December 6, 2005.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 Technical Specification
3.8.3.1, ``Onsite Power Distribution,'' to extend the allowed outage
time for balance-of-plant vital inverters 1-EDE-I-1E and 1-EDE-I-1F
from 24 hours to 7 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change extends the allowed outage time (AOT) for
the balance-of-plant (BOP) instrument bus inverters from 24 hours to
7 days. The BOP instrument bus inverters do not solely support any
risk-significant functions. The failure of an inverter is not an
initiator of any analyzed event and does not increase the frequency
of an initiating event. Consequently, extending the AOT will not
have an impact on the frequency of occurrence of any event
previously analyzed. The proposed change does not alter the design,
configuration, operation, or function of any plant system,
structure, or component. As a result, the outcomes of previously
evaluated accidents are unaffected. Therefore, the proposed change
does 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 kind of accident from any accident previously
evaluated.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change does not challenge the performance or integrity
of any safety-related system. The proposed change neither installs
nor removes any plant equipment, not alters the design, physical
configuration, or mode of operation of any plant structure, system,
or component. Installed equipment will not be operated in a new or
different manner. No physical changes are being made to the plant,
so no new accident causal mechanisms are being introduced.
Procedures that ensure the unit operates within analyzed limits and
procedures that respond to off-normal and emergency conditions are
not altered with this proposed change. Therefore, the proposed
change does not create the possibility of a new or different
accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in [a] margin of safety.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change does not alter the
design, configuration, operation, or function of any plant system,
structure, or component. The ability of any operable structure,
system, or component to perform its designated safety function is
unaffected by this change. Operation with one instrument bus
inverter inoperable and the associated instrument bus aligned to its
maintenance supply does not result in a significant reduction in [a]
margin of safety. Surveillance testing of the emergency diesel
generators (EDGs) and the electrical distribution system provides
confidence that the EDGs will energize the emergency AC buses
following a loss of power. 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Darrell J. Roberts.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: November 12, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.7, ``Inservice Testing
Program,'' and TS 5.5.8, ``Steam Generator (SG) Tube Surveillance
Program,'' to update references to the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (Code) and certain
associated periodicities for inservice testing activities consistent
with the requirements of Title 10 of the Code of Federal Regulations
(10 CFR) section 50.55a, ``Codes and standards.''
[[Page 2593]]
The proposed amendment would also correct a typographical error
contained in TS 5.5.8.b.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
The proposed change revises Technical Specifications for
consistency with the requirements of 10 CFR 50.55a(f)(4) and 10 CFR
50.55a(g)(4).
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing pumps and
valves.
The proposed 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 actuation setpoints, accident mitigation capabilities, or
accident analysis assumptions or inputs.
Therefore, the probability or consequences of any accident
previously evaluated will not be significantly increased as a result
of the proposed change.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing. The
proposed change does not involve a modification to the physical
configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Equipment important to safety will continue to operate as
designed. The changes do not result in any event previously deemed
incredible been made credible. The changes do not result in adverse
conditions or result in any increase in the challenges to safety
systems. Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendment will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The proposed change incorporates revisions to the ASME Code that
result in a net improvement in the measures for testing. The safety
function of the affected components will be maintained.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other structures,
systems or components (SSCs) important to safety. Therefore, the
requested change will not result in 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 11, 2005.
Description of amendment request: The proposed amendment would
revise certain 18-month Technical Specification (TS) Surveillance
Requirements (SRs) to eliminate the condition that testing be conducted
during shutdown.
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 changes permit PSEG to evaluate the conditions
required to safely perform a TS SR. These surveillance tests verify
that equipment will perform its intended safety function of
mitigating an accident. No analyzed accident scenario is being
revised. The initiating conditions and assumptions for accidents
described in the Hope Creek Generating Station Updated Final Safety
Analysis Report (UFSAR) remain as previously analyzed.
The proposed changes do not reduce the ability of the mitigating
equipment to perform its safety function. The TS will continue to
require the surveillance tests to be performed on an eighteen-month
periodicity to verify operability. As a result, the ability of the
mitigating equipment to perform its safety function is unaffected by
the proposed change.
The capitalization change is proposed to improve readability and
does not alter any requirement.
Based upon the above, the proposed changes will not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
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 create the possibility of a new or
different kind of accident from any accident previously evaluated in
the UFSAR. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. Specifically, no new hardware is being added to the plant
as part of the proposed change, no existing equipment is being
modified, and no significant changes in operations are being
introduced.
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 will not alter any assumptions, initial
conditions, or results of any accident analyses. The proposed
changes to remove the requirement to perform certain testing during
shutdown conditions allows PSEG to evaluate the conditions needed to
safely perform the required testing. There is no change to the
frequency of testing or in the testing that is required. There is no
change in the responsibility of PSEG to perform tests in a safe and
responsible manner. Any changes to procedures will have to be
individually evaluated to ensure that they do not reduce the margin
of safety. The changes do not affect the ability of systems,
structures or components to perform their safety related functions.
In addition, the proposed changes do not affect the ability of the
safety systems to ensure that the facility can be maintained in a
shutdown or refueling condition for extended periods of time.
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 Branch Chief: Darrell J. Roberts.
[[Page 2594]]
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 amendment request: August 31, 2005; as supplemented
December 8, 2005.
Description of amendment request: The proposed amendment would
relocate the containment high range accident monitors from the
radiation monitoring instrumentation technical specification (TS) to
the accident monitoring TS and correct a typographical error contained
in a previous amendment.
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 presents no change in the probability of a
previously evaluated accident.
The proposed change presents no change in the consequence of an
accident, since the containment high range accident monitors are
used post-accident to determine the amount of core damage and status
of the fission product barriers.
The containment high range accident monitors are used post
accident to assess the conditions inside containment. They have an
automatic function to switch the subcooling margin monitor (SCMM) to
``adverse'' mode (i.e., it displays a more conservative indication
of the amount of subcooling in the RCS) [reactor coolant system].
Additionally, the containment high range accident monitors provide
an indication that is used post accident in determining the status
of the fission product barriers. There will be no change in the
operation or use of the containment high range accident monitors.
The remaining change is editorial in nature and does not impact
the accident analysis in any manner.
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 is a minor change that is administrative in
nature. No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
No new hardware is added, existing hardware is not modified and no
significant changes in operations are implemented. Post accident
monitoring instrumentation is not associated with the initiation of
an accident.
3. Does the proposed change involve a significant reduction in
[a] margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety systems settings or limiting conditions for
operation are determined. The proposed change will not alter any
assumptions, initial conditions or results specified in any accident
analysis.
There is no change in the containment high range accident
monitor high level alarm setpoint. The ECS [electronic check source]
is functionally equivalent to the TS definition of SOURCE CHECK.
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 Branch Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey