[Federal Register Volume 71, Number 161 (Monday, August 21, 2006)]
[Notices]
[Pages 48561-48564]
From the Federal Register Online via the Government Publishing 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 [email protected].
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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