Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement To Modify Requirements Regarding LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation Using the Consolidated Line Item Improvement Process, 48561-48564 [E6-13715]
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Federal Register / Vol. 71, No. 161 / Monday, August 21, 2006 / Notices
the license termination criteria in 10
CFR 20.1402, ‘‘Radiological Criteria for
Unrestricted Use.’’
The staff has examined the licensee’s
request and the information provided in
support of its request, including the
surveys performed to demonstrate
compliance with the release criteria.
The staff has found that the radiological
environmental impacts from the
proposed action are bounded by the
impacts evaluated in the ‘‘Generic
Environmental Impact Statement in
Support of Rulemaking on Radiological
Criteria for License Termination of NRCLicensed Facilities’’ (NUREG–1496).
Additionally, no non-radiological or
cumulative impacts were identified.
Based on its review, the staff has
determined that there are no additional
remediation activities necessary to
complete the proposed action and a
Finding of No Significant Impact is
appropriate.
III. Finding of No Significant Impact
On the basis of the EA, the NRC
concluded that there are no significant
environmental impacts from the
proposed amendment and determined
not to prepare an environmental impact
statement.
hsrobinson on PROD1PC72 with NOTICES
IV. Further Information
Documents related to this action,
including the application for
amendment and supporting
documentation, are available
electronically at the NRC’s electronic
Reading Room at https://www.nrc.gov/
reading-rm/adams.html. From this site,
you can access the NRC’s Agencywide
Document Access and Management
System (ADAMS), which provides text
and image files of NRC’s public
documents. The ADAMS accession
numbers for the documents related to
this notice are: ML060690446 for the
March 7, 2006, license termination
request, ML061980294 for the July 11,
2006, additional information to the
amendment request, and ML062190210
for the EA summarized above. If you do
not have access to ADAMS or if there
are problems in accessing the
documents located in ADAMS, contact
the NRC’s Public Document Room (PDR)
Reference staff at 1–800–397–4209, 301–
415–4737, or by e-mail to pdr@nrc.gov.
These documents may also be viewed
electronically on the public computers
located at the NRC’s PDR, O 1 F21, One
White Flint North, 11555 Rockville
Pike, Rockville, MD 20852. The PDR
reproduction contractor will copy
documents for a fee.
Dated at Lisle, Illinois, this 10th day of
August 2006.
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For the Nuclear Regulatory Commission.
Jamnes L. Cameron,
Chief, Decommissioning Branch, Division of
Nuclear Materials Safety, Region III.
[FR Doc. E6–13718 Filed 8–18–06; 8:45 am]
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48561
received may be examined at the NRC’s
Public Document Room, 11555
Rockville Pike (Room O–1F21),
Rockville, Maryland. Comments may be
submitted by electronic mail to
NRCREP@nrc.gov.
Tim
Kobetz, Mail Stop: O–12H2, Division of
Inspections and Regional Support,
Office of Nuclear Reactor Regulation,
U.S. Nuclear Regulatory Commission,
Washington, DC 20555–0001, telephone
301–415–1932.
FOR FURTHER INFORMATION CONTACT:
NUCLEAR REGULATORY
COMMISSION
Notice of Opportunity To Comment on
Model Safety Evaluation on Technical
Specification Improvement To Modify
Requirements Regarding LCO 3.10.1,
Inservice Leak and Hydrostatic Testing
Operation Using the Consolidated Line
Item Improvement Process
Nuclear Regulatory
Commission.
ACTION: Request for comment.
AGENCY:
SUMMARY: Notice is hereby given that
the staff of the Nuclear Regulatory
Commission (NRC) has prepared a
model safety evaluation (SE) relating to
the modification of shutdown testing
requirements in technical specifications
(TS) for Boiling Water Reactors (BWR).
The NRC staff has also prepared a model
no-significant-hazards-consideration
(NSHC) determination relating to this
matter. The purpose of these models is
to permit the NRC to efficiently process
amendments that propose to modify
LCO 3.10.1 that would allow control rod
scram time testing to be performed
concurrently with inservice leak and
hydrostatic testing. Licensees of nuclear
power reactors to which the models
apply could then request amendments,
confirming the applicability of the SE
and NSHC determination to their
reactors. The NRC staff is requesting
comment on the model SE and model
NSHC determination prior to
announcing their availability for
referencing in license amendment
applications.
DATES: The comment period expires
September 20, 2006. Comments received
after this date will be considered if it is
practical to do so, but the Commission
is able to ensure consideration only for
comments received on or before this
date.
ADDRESSES: Comments may be
submitted either electronically or via
U.S. mail. Submit written comments to
Chief, Rules and Directives Branch,
Division of Administrative Services,
Office of Administration, Mail Stop: T–
6 D59, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001. Hand deliver comments to: 11545
Rockville Pike, Rockville, Maryland,
between 7:45 a.m. and 4:15 p.m. on
Federal workdays. Copies of comments
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SUPPLEMENTARY INFORMATION:
Background
Regulatory Issue Summary 2000–06,
‘‘Consolidated Line Item Improvement
Process for Adopting Standard
Technical Specification Changes for
Power Reactors,’’ was issued on March
20, 2000. The consolidated line item
improvement process (CLIIP) is
intended to improve the efficiency of
NRC licensing processes by processing
proposed changes to the standard
technical specifications (STS) in a
manner that supports subsequent
license amendment applications. The
CLIIP includes an opportunity for the
public to comment on a proposed
change to the STS after a preliminary
assessment by the NRC staff and a
finding that the change will likely be
offered for adoption by licensees. This
notice solicits comment on a proposal to
modify LCO 3.10.1 that would allow
control rod scram time testing to be
performed concurrently with inservice
leak and hydrostatic testing. The CLIIP
directs the NRC staff to evaluate any
comments received for a proposed
change to the STS and to either
reconsider the change or announce the
availability of the change for adoption
by licensees.
This notice involves the modification
of LCO 3.10.1 that would allow control
rod scram time testing to be performed
concurrently with inservice leak and
hydrostatic testing. This change was
proposed for incorporation into the
standard technical specifications by the
owners groups participants in the
Technical Specification Task Force
(TSTF) and is designated TSTF–484.
TSTF–484 can be viewed on the NRC’s
Web page utilizing the Agencywide
Documents Access and Management
System (ADAMS). ADAMS accession
numbers are ML052930102 (TSTF–484
Submittal), ML060970568 (NRC Request
for Additional Information, RAI), and
ML061560523 (TSTF Response to NRC
RAIs).
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Federal Register / Vol. 71, No. 161 / Monday, August 21, 2006 / Notices
Applicability
Licensees opting to apply for this TS
change are responsible for reviewing the
staff’s evaluation, referencing the
applicable technical justifications, and
providing any necessary plant-specific
information. Each amendment
application made in response to the
notice of availability will be processed
and noticed in accordance with
applicable rules and NRC procedures.
Public Notices
This notice requests comments from
interested members of the public within
30 days of the date of publication in the
Federal Register. After evaluating the
comments received as a result of this
notice, the staff will either reconsider
the proposed change or announce the
availability of the change in a
subsequent notice (perhaps with some
changes to the safety evaluation or the
proposed no significant hazards
consideration determination as a result
of public comments). If the staff
announces the availability of the
change, licensees wishing to adopt the
change must submit an application in
accordance with applicable rules and
other regulatory requirements. For each
application the staff will publish a
notice of consideration of issuance of
amendment to facility operating
licenses, a proposed no significant
hazards consideration determination,
and a notice of opportunity for a
hearing. The staff will also publish a
notice of issuance of an amendment to
an operating license to announce the
modification of TS 3.10.1, Inservice
Leak and Hydrostatic Testing, for each
plant that receives the requested change.
hsrobinson on PROD1PC72 with NOTICES
Proposed Safety Evaluation—U.S.
Nuclear Regulatory Commission, Office
of Nuclear Reactor Regulation,
Consolidated Line Item Improvement,
Technical Specification Task Force
(TSTF) Change TSTF–484, Revision 0,
Use of TS 3.10.1 for Scram Time
Testing Activities
1.0 Introduction
By application dated [Date], [Name of
Licensee] (the licensee) requested
changes to the Technical Specifications
(TS) for the [Name of Facility].
The proposed changes would revise
LCO 3.10.1, and the associated Bases, to
expand its scope to include provisions
for temperature excursions greater than
[200]°F as a consequence of inservice
leak and hydrostatic testing, and as a
consequence of scram time testing
initiated in conjunction with an
inservice leak or hydrostatic test, while
considering operational conditions to be
in Mode 4.
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2.0
Regulatory Evaluation
2.1 Inservice Leak and Hydrostatic
Testing
The Reactor Coolant System (RCS)
serves as a pressure boundary and also
serves to provide a flow path for the
circulation of coolant past the fuel. In
order to maintain RCS integrity, Section
XI of the American Society of
Mechanical Engineers (ASME) Pressure
Vessel Code requires periodic
hydrostatic and leakage testing.
Hydrostatic tests are required to be
performed once every 10 years and
Leakage tests are required to be
performed each refueling outage.
Appendix G to 10 CFR Part 50 states
that pressure tests and leak tests of the
reactor vessel that are required by
Section XI of the American Society of
Mechanical Engineers (ASME) Pressure
Vessel Code must be completed before
the core is critical.
NUREG–1433, General Electric Plants,
BWR/4, Revision 3, Standard Technical
Specifications (STS) and NUREG–1434,
General Electric Plants, BWR/6,
Revision 3, STS both currently contain
LCO 3.10.1, Inservice Leak and
Hydrostatic Testing Operation. LCO
3.10.1 was created to allow for
hydrostatic and leakage testing to be
conducted while in Mode 4 with
average reactor coolant temperature
greater than [200]°F provided certain
secondary containment LCOs are met.
TSTF–484, Revision 0, Use of TS
3.10.1 for Scram Time Testing
Activities, modifies LCO 3.10.1 to allow
a licensee to implement LCO 3.10.1
while hydrostatic and leakage testing is
being conducted should average reactor
coolant temperature exceed [200]°F
during testing. This modification does
not alter current requirements for
hydrostatic and leakage testing as
required by Appendix G to 10 CFR part
50.
2.2 Control Rod Scram Time Testing
Control Rods function to control
reactor power level and to provide
adequate excess negative reactivity to
shut down the reactor from any normal
operating or accident condition at any
time during core life. The control rods
are scrammed by using hydraulic
pressure exerted by the Control Rod
Drive (CRD) system. Criterion 10 of
Appendix A to 10 CFR part 50 states
that the reactor core and associated
coolant, control, and protection systems
shall be designed with appropriate
margin to assure that specified
acceptable fuel limits are not exceed
during any condition of normal
operation, including the effects of
anticipated operational occurrences.
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The scram reactivity used in design
basis accidents (DBA) and transient
analyses is based on an assumed control
rod scram time.
NUREG–1433, General Electric Plants,
BWR/4, Revision 3, Standard Technical
Specifications (STS) and NUREG–1434,
General Electric Plants, BWR/6,
Revision 3, STS both currently contain
surveillance requirements (SR) to
conduct scram time testing when certain
conditions are met in order to ensure
that Criterion 10 of Appendix A to 10
CFR part 50 is satisfied. SR 3.1.4.1
requires scram time testing to be
conducted following a shutdown greater
than 120 days while SR 3.1.4.4 requires
scram time testing to be conducted
following work on the CRD system or
following fuel movement within the
affected core cell. Both SR must be
performed at reactor pressure greater
than or equal to [800] psig and prior to
initially exceeding 40% rated thermal
power (RTP).
TSTF–484, Revision 0, Use of TS
3.10.1 for Scram Time Testing
Activities, would modify LCO 3.10.1 to
allow SR 3.1.4.1 and SR 3.1.4.4 to be
conducted in Mode 4 with average
reactor coolant temperature greater than
[200]°F. Scram time testing would be
performed in accordance with LCO
3.10.4, Single Control Rod
Withdrawal—Cold Shutdown. This
modification to LCO 3.10.1 does not
alter the means of compliance with
Criterion 10 of Appendix A to 10 CFR
part 50.
3.0 Technical Evaluation
The existing provisions of LCO 3.10.1
allow for hydrostatic and leakage testing
to be conducted while in Mode 4 with
average reactor coolant temperature
greater than [200]°F, while imposing
Mode 3 secondary containment
requirements. Under the existing
provision, LCO 3.10.1 would have to be
implemented prior to hydrostatic and
leakage testing. As a result, if LCO
3.10.1 was not implemented prior to
hydrostatic and leakage testing,
hydrostatic and leakage testing would
have to be terminated if average reactor
coolant temperature exceeded [200]°F
during the conduct of the hydrostatic
and leakage test. TSTF–484, Revision 0,
Use of TS 3.10.1 for Scram Time Testing
Activities, modifies LCO 3.10.1 to allow
a licensee to implement LCO 3.10.1
while hydrostatic and leakage testing is
being conducted should average reactor
coolant temperature exceed [200]°F
during testing. The modification will
allow completion of testing without the
potential for interrupting the test in
order to reduce reactor vessel pressure,
cool the RCS, and restart the test below
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Federal Register / Vol. 71, No. 161 / Monday, August 21, 2006 / Notices
[200]°F. Since the current LCO 3.10.1
allows testing to be conducted while in
Mode 4 with average reactor coolant
temperature greater than [200]°F, the
proposed change does not introduce any
new operational conditions beyond
those currently allowed.
Surveillance Requirements (SR)
3.1.4.1 and SR 3.1.4.4 require that
control rod scram time be tested at
reactor pressure greater than or equal to
[800] psig and before exceeding 40%
rated thermal power (RTP). Performance
of control rod scram time testing is
typically scheduled concurrent with
inservice leak or hydrostatic testing
while the reactor coolant system (RCS)
is pressurized. Because of the number of
control rods that must be tested, it is
possible for the inservice leak or
hydrostatic test to be completed prior to
completing the scram time test. Under
existing provisions, if scram time testing
can not be completed during the LCO
3.10.1 inservice leak or hydrostatic test,
scram time testing must be suspended.
Additionally, if LCO 3.10.1 is not
implemented and average reactor
coolant temperature exceeds [200]°F
while performing the scram time test,
scram time testing must also be
suspended. In both situations, scram
time testing is resumed during startup
prior to exceeding 40% RTP. TSTF–484,
Revision 0, Use of TS 3.10.1 for Scram
Time Testing Activities, modifies LCO
3.10.1 to allow a licensee to complete
scram time testing initiated during
inservice leak or hydrostatic testing. As
stated earlier, since the current LCO
3.10.1 allows testing to be conducted
while in Mode 4 with average reactor
coolant temperature greater than
[200]°F, the proposed change does not
introduce any new operational
conditions beyond those currently
allowed. Completion of scram time
testing prior to reactor criticality and
power operations results in a more
conservative operating philosophy with
attendant potential safety benefits.
It is acceptable to perform other
testing concurrent with the inservice
leak or hydrostatic test provided that
this testing can be performed safely and
does not interfere with the leak or
hydrostatic test. However, it is not
permissible to remain in TS 3.10.1
solely to complete such testing
following the completion of inservice
leak or hydrostatic testing and scram
time testing.
Since the tests are performed with the
reactor pressure vessel (RPV) nearly
water solid, at low decay heat values,
and near Mode 4 conditions, the stored
energy in the reactor core will be very
low. Small leaks from the RCS would be
detected by inspections before a
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significant loss of inventory occurred. In
addition, two low pressure emergency
core cooling systems (ECCS) injection/
spray subsystems are required to be
operable in Mode 4 by TS 3.5.2, ECCSShutdown. In the event of a large RCS
leak, the RPV would rapidly
depressurize and allow operation of the
low pressure ECCS. The capability of
the low pressure ECCS would be
adequate to maintain the fuel covered
under the low decay heat conditions
during these tests. Also, LCO 3.10.1
requires that secondary containment
and standby gas treatment system be
operable and capable of handling any
airborne radioactivity or steam leaks
that may occur during performance of
testing.
The protection provided by the
normally required Mode 4 applicable
LCOs, in addition to the secondary
containment requirements required to
be met by LCO 3.10.1, minimizes
potential consequences in the event of
any postulated abnormal event during
testing. In addition, the requested
modification to LCO 3.10.1 does not
create any new modes of operation or
operating conditions that are not
currently allowed.
4.0 State Consultation
In accordance with the Commission’s
regulations, the [Name of State] State
official was notified of the proposed
issuance of the amendment. The State
official had [no] comments. [If
comments were provided, they should
be addressed here].
5.0 Environmental Consideration
The amendment changes a
requirement with respect to installation
or use of a facility component located
within the restricted area as defined in
10 CFR part 20. The NRC staff has
determined that the amendment
involves no significant increase in the
amounts, and no significant change in
the types, of any effluents that may be
released offsite, and that there is no
significant increase in individual or
cumulative occupational radiation
exposure. A significant hazards
consideration is attached and is
available for public comment. The
amendment meets the eligibility criteria
for categorical exclusion set forth in 10
CFR 51.22(c)(9). Pursuant to 10 CFR
51.22(b) no environmental impact
statement or environmental assessment
need be prepared in connection with the
issuance of the amendment.
6.0 Conclusion
The Commission has concluded,
based on the considerations discussed
above, that: (1) There is reasonable
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48563
assurance that the health and safety of
the public will not be endangered by
operation in the proposed manner, (2)
such activities will be conducted in
compliance with the Commission’s
regulations, and (3) the issuance of the
amendments will not be inimical to the
common defense and security or to the
health and safety of the public.
7.0
References
1. NUREG–1433, ‘‘General Electric Plants,
BWR/4, Revision 3, Standard Technical
Specifications (STS)’’, August 31, 2003.
2. NUREG–1434, General Electric Plants,
BWR/6, Revision 3, Standard Technical
Specifications (STS)’’, August 31, 2003.
3. Request for Additional Information (RAI)
Regarding TSTF–484, April, 7, 2006, ADAMS
accession number ML060970568.
4. Response to NRC RAIs Regarding TSTF–
484, June 5, 2006, ADAMS accession number
ML061560523.
5. TSTF–484 Revision 0, ‘‘Use of TS 3.10.1
for Scram Times Testing Activities’’, May 5,
2005, ADAMS accession number
ML052930102.
Model No Significant Hazards
Determination
Description of Amendment Request:
The proposed changes would revise
LCO 3.10.1, and the associated Bases, to
expand its scope to include provisions
for temperature excursions greater than
[200]°F as a consequence of inservice
leak and hydrostatic testing, and as a
consequence of scram time testing
initiated in conjunction with an
inservice leak or hydrostatic test, while
considering operational conditions to be
in Mode 4.
Basis for No Significant Hazards
Determination: As required by 10 CFR
50.91 (a), an analysis of the issue of no
significant hazards consideration is
presented below:
Criterion 1: The proposed change does
not involve a significant increase in the
probability or consequences of an
accident previously evaluated.
Technical Specifications currently
allow for operation at greater than
[200]°F while imposing MODE 4
requirements in addition to the
secondary containment requirements
required to be met. Extending the
activities that can apply this allowance
will not adversely impact the
probability or consequences of an
accident previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an
accident previously evaluated.
Criterion 2: The proposed change does
not create the possibility of a new or
different kind of accident from any
accident previously evaluated.
Technical Specifications currently
allow for operation at greater than
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Federal Register / Vol. 71, No. 161 / Monday, August 21, 2006 / Notices
hsrobinson on PROD1PC72 with NOTICES
[200]°F while imposing MODE 4
requirements in addition to the
secondary containment requirements
required to be met. No new operational
conditions beyond those currently
allowed by LCO 3.10.1 are introduced.
The changes do not involve a physical
alteration of the plant (i.e., no new or
different type of equipment will be
installed) or a change in the methods
governing normal plant operation. In
addition, the changes do not impose any
new or different requirements or
eliminate any existing requirements.
The changes do not alter assumptions
made in the safety analysis. The
proposed changes are consistent with
the safety analysis assumptions and
current plant operating practice.
Therefore, the proposed change does not
create the possibility of a new or
different kind of accident from any
accident previously evaluated.
Criterion 3: The proposed change does
not involve a significant reduction in a
margin of safety.
Technical Specifications currently
allow for operation at greater than
[200]°F while imposing MODE 4
requirements in addition to the
secondary containment requirements
required to be met. Extending the
activities that can apply this allowance
will not adversely impact any margin of
safety. Allowing completion of
inspections and testing and supporting
completion of scram time testing
initiated in conjunction with an
inservice leak or hydrostatic test prior to
power operation results in enhanced
safe operations by eliminating
unnecessary maneuvers to control
reactor temperature and pressure.
Therefore, the proposed change does not
involve a significant reduction in a
margin of safety.
Based on the above, the NRC
concludes that the proposed change
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.
Principal Contributor: Aron Lewin.
Dated at Rockville, Maryland this 15th day
of August 2006.
For the Nuclear Regulatory Commission.
Timothy Kobetz,
Branch Chief, Technical Specifications
Branch, Division of Inspections and Regional
Support, Office of Nuclear Reactor
Regulation.
[FR Doc. E6–13715 Filed 8–18–06; 8:45 am]
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NUCLEAR WASTE TECHNICAL
REVIEW BOARD
Notice of a Meeting; Yucca Mountain,
NV
Workshop: September 25–26, 2006—
Las Vegas, Nevada; The U.S. Nuclear
Waste Technical Review board will host
a workshop on the potential for
localized corrosion of Alloy-22, the
material that has been proposed for
waste packages in which spent nuclear
fuel and high-level radioactive waste
will be disposed of inside the proposed
Yucca Mountain repository.
Pursuant to its authority under
section 5051 of Public Law 100–203,
Nuclear Waste Policy Amendments Act
of 1987, the U.S. Nuclear Waste
Technical Review Board will host a
workshop on localized corrosion in Las
Vegas, Nevada. The focus of the
workshop will be the potential for
localized corrosion of Alloy-22 under
aqueous conditions that might exist in
a proposed Yucca Mountain repository.
Alloy-22 is a material that has been
proposed for waste packages in which
spent nuclear fuel and high-level
radioactive waste will be disposed of
inside the proposed repository. Among
the workshop topics will be results of
recent and ongoing testing related to
evolution of aqueous environments in
the repository and the potential
initiation, propagation, cessation, and
consequences of localized corrosion of
Alloy-22. The Board was charged in the
Nuclear Waste Amendments Act of 1987
with conducting an independent review
of the technical and scientific validity of
U.S. Department of Energy (DOE)
activities related to disposing,
packaging, and transporting of spent
nuclear fuel and high-level radioactive
waste.
The workshop agenda will be
available on the Board’s Web site
https://www.nwtrb.gov) approximately
one week before the date of the
workshop. The agenda also may be
obtained by telephone request at that
time. The workshop will be open to the
public, and opportunities for public
comment will be provided. Transcripts
of the workshop proceedings and
overheads from workshop presentations
will be available on the Board’s Web site
approximately three weeks after the
workshop date.
The workshop will be held at the Las
Vegas Marriott Suites; 325 Convention
Center Drive; Las Vegas, Nevada 89109;
telephone 702–650–2000; fax 702–650–
9466.
The workshop will begin Monday
afternoon with introductions of the
participants; presentations of the ground
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rules; and a discussion of possible waste
package environments, including data
obtained from current and ongoing tests,
interpretation of the data, and modeling
used to project possible waste package
environments.
On Tuesday morning, the workshop
will reconvene, and discussions will
focus on testing related to the potential
for localized corrosion of the Alloy-22
waste packages. The discussions will
continue until late afternoon, when the
workshop will adjourn.
Time will be set aside during the
workshop for public comments. Those
wanting to speak are encouraged to sign
the ‘‘Public Comment Register’’ at the
check-in-table. A time limit may have to
be set on individual remarks, but
written comments of any length may be
submitted for the record.
Transcripts of the workshop will be
available on the Board’s Web site, by email, on computer disk, and on a
library-loan basis in paper format from
Davonya Barnes of the Board’s staff no
later than October 19, 2006.
A block of rooms has been reserved
for workshop attendees and participants
at the Las Vegas Marriott Suites. When
making a reservation, please state that
you will be attending the Nuclear Waste
Technical Review Board workshop.
Reservations should be made by
September 1, 2006, to ensure receiving
the workshop rate.
For more information, contact Karyn
Severson, NWTRB External Affairs;
2300 Clarendon Boulevard, Suite 1300;
Arlington, VA 22201–3367; 703–235–
4473; fax 703–235–4495.
Dated: August 16, 2006.
William D. Barnard,
Executive Director, Nuclear Waste Technical
Review Board.
[FR Doc. 06–7049 Filed 8–18–06; 8:45am]
BILLING CODE 6820–AM–M
NUCLEAR WASTE TECHNICAL
REVIEW BOARD
Notice of a Board Meeting; Amargosa
Valley, NV
Board meeting: September 27, 2006—
Amargosa Valley, Nevada; The U.S.
Nuclear Waste Technical Review Board
will meet to discuss U.S. Department of
Energy efforts to develop and articulate
a safety case for the proposed Yucca
Mountain repository.
Pursuant to its authority under
section 5051 of Public Law 100–203,
Nuclear Waste Policy Amendments Act
of 1987, the U.S. Nuclear Waste
Technical Review Board will meet in
Amargosa Valley, Nevada, on
Wednesday, September 27, 2006, to
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]]>Agencies
[Federal Register Volume 71, Number 161 (Monday, August 21, 2006)]
[Notices]
[Pages 48561-48564]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-13715]
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NUCLEAR REGULATORY COMMISSION
Notice of Opportunity To Comment on Model Safety Evaluation on
Technical Specification Improvement To Modify Requirements Regarding
LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation Using the
Consolidated Line Item Improvement Process
AGENCY: Nuclear Regulatory Commission.
ACTION: Request for comment.
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SUMMARY: Notice is hereby given that the staff of the Nuclear
Regulatory Commission (NRC) has prepared a model safety evaluation (SE)
relating to the modification of shutdown testing requirements in
technical specifications (TS) for Boiling Water Reactors (BWR). The NRC
staff has also prepared a model no-significant-hazards-consideration
(NSHC) determination relating to this matter. The purpose of these
models is to permit the NRC to efficiently process amendments that
propose to modify LCO 3.10.1 that would allow control rod scram time
testing to be performed concurrently with inservice leak and
hydrostatic testing. Licensees of nuclear power reactors to which the
models apply could then request amendments, confirming the
applicability of the SE and NSHC determination to their reactors. The
NRC staff is requesting comment on the model SE and model NSHC
determination prior to announcing their availability for referencing in
license amendment applications.
DATES: The comment period expires September 20, 2006. Comments received
after this date will be considered if it is practical to do so, but the
Commission is able to ensure consideration only for comments received
on or before this date.
ADDRESSES: Comments may be submitted either electronically or via U.S.
mail. Submit written comments to Chief, Rules and Directives Branch,
Division of Administrative Services, Office of Administration, Mail
Stop: T-6 D59, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001. Hand deliver comments to: 11545 Rockville Pike, Rockville,
Maryland, between 7:45 a.m. and 4:15 p.m. on Federal workdays. Copies
of comments received may be examined at the NRC's Public Document Room,
11555 Rockville Pike (Room O-1F21), Rockville, Maryland. Comments may
be submitted by electronic mail to NRCREP@nrc.gov.
FOR FURTHER INFORMATION CONTACT: Tim Kobetz, Mail Stop: O-12H2,
Division of Inspections and Regional Support, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, telephone 301-415-1932.
SUPPLEMENTARY INFORMATION:
Background
Regulatory Issue Summary 2000-06, ``Consolidated Line Item
Improvement Process for Adopting Standard Technical Specification
Changes for Power Reactors,'' was issued on March 20, 2000. The
consolidated line item improvement process (CLIIP) is intended to
improve the efficiency of NRC licensing processes by processing
proposed changes to the standard technical specifications (STS) in a
manner that supports subsequent license amendment applications. The
CLIIP includes an opportunity for the public to comment on a proposed
change to the STS after a preliminary assessment by the NRC staff and a
finding that the change will likely be offered for adoption by
licensees. This notice solicits comment on a proposal to modify LCO
3.10.1 that would allow control rod scram time testing to be performed
concurrently with inservice leak and hydrostatic testing. The CLIIP
directs the NRC staff to evaluate any comments received for a proposed
change to the STS and to either reconsider the change or announce the
availability of the change for adoption by licensees.
This notice involves the modification of LCO 3.10.1 that would
allow control rod scram time testing to be performed concurrently with
inservice leak and hydrostatic testing. This change was proposed for
incorporation into the standard technical specifications by the owners
groups participants in the Technical Specification Task Force (TSTF)
and is designated TSTF-484. TSTF-484 can be viewed on the NRC's Web
page utilizing the Agencywide Documents Access and Management System
(ADAMS). ADAMS accession numbers are ML052930102 (TSTF-484 Submittal),
ML060970568 (NRC Request for Additional Information, RAI), and
ML061560523 (TSTF Response to NRC RAIs).
[[Page 48562]]
Applicability
Licensees opting to apply for this TS change are responsible for
reviewing the staff's evaluation, referencing the applicable technical
justifications, and providing any necessary plant-specific information.
Each amendment application made in response to the notice of
availability will be processed and noticed in accordance with
applicable rules and NRC procedures.
Public Notices
This notice requests comments from interested members of the public
within 30 days of the date of publication in the Federal Register.
After evaluating the comments received as a result of this notice, the
staff will either reconsider the proposed change or announce the
availability of the change in a subsequent notice (perhaps with some
changes to the safety evaluation or the proposed no significant hazards
consideration determination as a result of public comments). If the
staff announces the availability of the change, licensees wishing to
adopt the change must submit an application in accordance with
applicable rules and other regulatory requirements. For each
application the staff will publish a notice of consideration of
issuance of amendment to facility operating licenses, a proposed no
significant hazards consideration determination, and a notice of
opportunity for a hearing. The staff will also publish a notice of
issuance of an amendment to an operating license to announce the
modification of TS 3.10.1, Inservice Leak and Hydrostatic Testing, for
each plant that receives the requested change.
Proposed Safety Evaluation--U.S. Nuclear Regulatory Commission, Office
of Nuclear Reactor Regulation, Consolidated Line Item Improvement,
Technical Specification Task Force (TSTF) Change TSTF-484, Revision 0,
Use of TS 3.10.1 for Scram Time Testing Activities
1.0 Introduction
By application dated [Date], [Name of Licensee] (the licensee)
requested changes to the Technical Specifications (TS) for the [Name of
Facility].
The proposed changes would revise LCO 3.10.1, and the associated
Bases, to expand its scope to include provisions for temperature
excursions greater than [200][deg]F as a consequence of inservice leak
and hydrostatic testing, and as a consequence of scram time testing
initiated in conjunction with an inservice leak or hydrostatic test,
while considering operational conditions to be in Mode 4.
2.0 Regulatory Evaluation
2.1 Inservice Leak and Hydrostatic Testing
The Reactor Coolant System (RCS) serves as a pressure boundary and
also serves to provide a flow path for the circulation of coolant past
the fuel. In order to maintain RCS integrity, Section XI of the
American Society of Mechanical Engineers (ASME) Pressure Vessel Code
requires periodic hydrostatic and leakage testing. Hydrostatic tests
are required to be performed once every 10 years and Leakage tests are
required to be performed each refueling outage. Appendix G to 10 CFR
Part 50 states that pressure tests and leak tests of the reactor vessel
that are required by Section XI of the American Society of Mechanical
Engineers (ASME) Pressure Vessel Code must be completed before the core
is critical.
NUREG-1433, General Electric Plants, BWR/4, Revision 3, Standard
Technical Specifications (STS) and NUREG-1434, General Electric Plants,
BWR/6, Revision 3, STS both currently contain LCO 3.10.1, Inservice
Leak and Hydrostatic Testing Operation. LCO 3.10.1 was created to allow
for hydrostatic and leakage testing to be conducted while in Mode 4
with average reactor coolant temperature greater than [200][deg]F
provided certain secondary containment LCOs are met.
TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing
Activities, modifies LCO 3.10.1 to allow a licensee to implement LCO
3.10.1 while hydrostatic and leakage testing is being conducted should
average reactor coolant temperature exceed [200][deg]F during testing.
This modification does not alter current requirements for hydrostatic
and leakage testing as required by Appendix G to 10 CFR part 50.
2.2 Control Rod Scram Time Testing
Control Rods function to control reactor power level and to provide
adequate excess negative reactivity to shut down the reactor from any
normal operating or accident condition at any time during core life.
The control rods are scrammed by using hydraulic pressure exerted by
the Control Rod Drive (CRD) system. Criterion 10 of Appendix A to 10
CFR part 50 states that the reactor core and associated coolant,
control, and protection systems shall be designed with appropriate
margin to assure that specified acceptable fuel limits are not exceed
during any condition of normal operation, including the effects of
anticipated operational occurrences. The scram reactivity used in
design basis accidents (DBA) and transient analyses is based on an
assumed control rod scram time.
NUREG-1433, General Electric Plants, BWR/4, Revision 3, Standard
Technical Specifications (STS) and NUREG-1434, General Electric Plants,
BWR/6, Revision 3, STS both currently contain surveillance requirements
(SR) to conduct scram time testing when certain conditions are met in
order to ensure that Criterion 10 of Appendix A to 10 CFR part 50 is
satisfied. SR 3.1.4.1 requires scram time testing to be conducted
following a shutdown greater than 120 days while SR 3.1.4.4 requires
scram time testing to be conducted following work on the CRD system or
following fuel movement within the affected core cell. Both SR must be
performed at reactor pressure greater than or equal to [800] psig and
prior to initially exceeding 40% rated thermal power (RTP).
TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing
Activities, would modify LCO 3.10.1 to allow SR 3.1.4.1 and SR 3.1.4.4
to be conducted in Mode 4 with average reactor coolant temperature
greater than [200][deg]F. Scram time testing would be performed in
accordance with LCO 3.10.4, Single Control Rod Withdrawal--Cold
Shutdown. This modification to LCO 3.10.1 does not alter the means of
compliance with Criterion 10 of Appendix A to 10 CFR part 50.
3.0 Technical Evaluation
The existing provisions of LCO 3.10.1 allow for hydrostatic and
leakage testing to be conducted while in Mode 4 with average reactor
coolant temperature greater than [200][deg]F, while imposing Mode 3
secondary containment requirements. Under the existing provision, LCO
3.10.1 would have to be implemented prior to hydrostatic and leakage
testing. As a result, if LCO 3.10.1 was not implemented prior to
hydrostatic and leakage testing, hydrostatic and leakage testing would
have to be terminated if average reactor coolant temperature exceeded
[200][deg]F during the conduct of the hydrostatic and leakage test.
TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing
Activities, modifies LCO 3.10.1 to allow a licensee to implement LCO
3.10.1 while hydrostatic and leakage testing is being conducted should
average reactor coolant temperature exceed [200][deg]F during testing.
The modification will allow completion of testing without the potential
for interrupting the test in order to reduce reactor vessel pressure,
cool the RCS, and restart the test below
[[Page 48563]]
[200][deg]F. Since the current LCO 3.10.1 allows testing to be
conducted while in Mode 4 with average reactor coolant temperature
greater than [200][deg]F, the proposed change does not introduce any
new operational conditions beyond those currently allowed.
Surveillance Requirements (SR) 3.1.4.1 and SR 3.1.4.4 require that
control rod scram time be tested at reactor pressure greater than or
equal to [800] psig and before exceeding 40% rated thermal power (RTP).
Performance of control rod scram time testing is typically scheduled
concurrent with inservice leak or hydrostatic testing while the reactor
coolant system (RCS) is pressurized. Because of the number of control
rods that must be tested, it is possible for the inservice leak or
hydrostatic test to be completed prior to completing the scram time
test. Under existing provisions, if scram time testing can not be
completed during the LCO 3.10.1 inservice leak or hydrostatic test,
scram time testing must be suspended. Additionally, if LCO 3.10.1 is
not implemented and average reactor coolant temperature exceeds
[200][deg]F while performing the scram time test, scram time testing
must also be suspended. In both situations, scram time testing is
resumed during startup prior to exceeding 40% RTP. TSTF-484, Revision
0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO
3.10.1 to allow a licensee to complete scram time testing initiated
during inservice leak or hydrostatic testing. As stated earlier, since
the current LCO 3.10.1 allows testing to be conducted while in Mode 4
with average reactor coolant temperature greater than [200][deg]F, the
proposed change does not introduce any new operational conditions
beyond those currently allowed. Completion of scram time testing prior
to reactor criticality and power operations results in a more
conservative operating philosophy with attendant potential safety
benefits.
It is acceptable to perform other testing concurrent with the
inservice leak or hydrostatic test provided that this testing can be
performed safely and does not interfere with the leak or hydrostatic
test. However, it is not permissible to remain in TS 3.10.1 solely to
complete such testing following the completion of inservice leak or
hydrostatic testing and scram time testing.
Since the tests are performed with the reactor pressure vessel
(RPV) nearly water solid, at low decay heat values, and near Mode 4
conditions, the stored energy in the reactor core will be very low.
Small leaks from the RCS would be detected by inspections before a
significant loss of inventory occurred. In addition, two low pressure
emergency core cooling systems (ECCS) injection/spray subsystems are
required to be operable in Mode 4 by TS 3.5.2, ECCS-Shutdown. In the
event of a large RCS leak, the RPV would rapidly depressurize and allow
operation of the low pressure ECCS. The capability of the low pressure
ECCS would be adequate to maintain the fuel covered under the low decay
heat conditions during these tests. Also, LCO 3.10.1 requires that
secondary containment and standby gas treatment system be operable and
capable of handling any airborne radioactivity or steam leaks that may
occur during performance of testing.
The protection provided by the normally required Mode 4 applicable
LCOs, in addition to the secondary containment requirements required to
be met by LCO 3.10.1, minimizes potential consequences in the event of
any postulated abnormal event during testing. In addition, the
requested modification to LCO 3.10.1 does not create any new modes of
operation or operating conditions that are not currently allowed.
4.0 State Consultation
In accordance with the Commission's regulations, the [Name of
State] State official was notified of the proposed issuance of the
amendment. The State official had [no] comments. [If comments were
provided, they should be addressed here].
5.0 Environmental Consideration
The amendment changes a requirement with respect to installation or
use of a facility component located within the restricted area as
defined in 10 CFR part 20. The NRC staff has determined that the
amendment involves no significant increase in the amounts, and no
significant change in the types, of any effluents that may be released
offsite, and that there is no significant increase in individual or
cumulative occupational radiation exposure. A significant hazards
consideration is attached and is available for public comment. The
amendment meets the eligibility criteria for categorical exclusion set
forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no
environmental impact statement or environmental assessment need be
prepared in connection with the issuance of the amendment.
6.0 Conclusion
The Commission has concluded, based on the considerations discussed
above, that: (1) There is reasonable assurance that the health and
safety of the public will not be endangered by operation in the
proposed manner, (2) such activities will be conducted in compliance
with the Commission's regulations, and (3) the issuance of the
amendments will not be inimical to the common defense and security or
to the health and safety of the public.
7.0 References
1. NUREG-1433, ``General Electric Plants, BWR/4, Revision 3,
Standard Technical Specifications (STS)'', August 31, 2003.
2. NUREG-1434, General Electric Plants, BWR/6, Revision 3,
Standard Technical Specifications (STS)'', August 31, 2003.
3. Request for Additional Information (RAI) Regarding TSTF-484,
April, 7, 2006, ADAMS accession number ML060970568.
4. Response to NRC RAIs Regarding TSTF-484, June 5, 2006, ADAMS
accession number ML061560523.
5. TSTF-484 Revision 0, ``Use of TS 3.10.1 for Scram Times
Testing Activities'', May 5, 2005, ADAMS accession number
ML052930102.
Model No Significant Hazards Determination
Description of Amendment Request: The proposed changes would revise
LCO 3.10.1, and the associated Bases, to expand its scope to include
provisions for temperature excursions greater than [200][deg]F as a
consequence of inservice leak and hydrostatic testing, and as a
consequence of scram time testing initiated in conjunction with an
inservice leak or hydrostatic test, while considering operational
conditions to be in Mode 4.
Basis for No Significant Hazards Determination: As required by 10
CFR 50.91 (a), an analysis of the issue of no significant hazards
consideration is presented below:
Criterion 1: The proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
Technical Specifications currently allow for operation at greater
than [200][deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact the
probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
Criterion 2: The proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Technical Specifications currently allow for operation at greater
than
[[Page 48564]]
[200][deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. No new
operational conditions beyond those currently allowed by LCO 3.10.1 are
introduced. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. In
addition, the changes do not impose any new or different requirements
or eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3: The proposed change does not involve a significant
reduction in a margin of safety.
Technical Specifications currently allow for operation at greater
than [200][deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact any
margin of safety. Allowing completion of inspections and testing and
supporting completion of scram time testing initiated in conjunction
with an inservice leak or hydrostatic test prior to power operation
results in enhanced safe operations by eliminating unnecessary
maneuvers to control reactor temperature and pressure. Therefore, the
proposed change does not involve a significant reduction in a margin of
safety.
Based on the above, the NRC concludes that the proposed change
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.
Principal Contributor: Aron Lewin.
Dated at Rockville, Maryland this 15th day of August 2006.
For the Nuclear Regulatory Commission.
Timothy Kobetz,
Branch Chief, Technical Specifications Branch, Division of Inspections
and Regional Support, Office of Nuclear Reactor Regulation.
[FR Doc. E6-13715 Filed 8-18-06; 8:45 am]
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