Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement To Revise Control Rod Notch Surveillance Frequency, Clarify SRM Insert Control Rod Action, and Clarify Frequency Example, 46103-46109 [E7-16138]
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Federal Register / Vol. 72, No. 158 / Thursday, August 16, 2007 / Notices
Error 2. On page N–4 of Appendix N,
The last paragraph discusses the
monetary value for collective dose
averted and discount rates that may be
used in ALARA calculations. In
particular, the paragraph includes the
following two sentences: ‘‘For doses
averted within the first 100 years, a
discount rate of 7% should be used. For
doses averted beyond 100 years, a 3%
discount rate should be used. ‘‘ The
discussion of discount rate in these two
sentences is incorrect. Therefore, these
two quoted sentences are withdrawn
from the guidance of NUREG–1757, Vol.
2 and should not be used.
Error 3. On page N–10 of Appendix N,
Table N.2 summarizes acceptable
parameter values for use in
decommissioning ALARA analyses.
This table includes a row describing the
monetary discount rate, r. Consistent
with Error 2, above, the description for
the second column (the ‘‘value’’
description) of the row on monetary
discount rate, r, is withdrawn from the
guidance of NUREG–1757, Vol. 2.
Error 4. On page N–12 of Appendix N,
Example 3 is an ALARA calculation for
removing surface soil contaminated
with a long-lived radionuclide. Use of
the single discount rate in the example
may be misleading, because the
guidance in NUREG/BR–0058
recommends multiple analyses be
performed. Therefore, Example 3 is
withdrawn from Appendix N of
NUREG–1757, Vol. 2, and should not be
used.
Error 5. On page N–18 of Appendix N,
the last paragraph again discusses
acceptable values for the discount rate,
r. In particular, this paragraph includes
the sentence: ‘‘Values for r are given in
NUREG/BR–0058, Revision 2, and OMB
policy (OMB 1996).’’ The referenced
guidance is out-of-date, and this quoted
sentence is withdrawn from the
guidance of NUREG–1757, Vol. 2.
The staff intends to develop interim
guidance to address the withdrawn
portions of guidance discussed above
and will post the interim guidance on
the NRC’s decommissioning Web page,
to make it available for use by licensees
and other stakeholders.
The guidance in NUREG–1757 and
any corrections to NUREG–1757 are
intended for use by NRC staff and
licensees. The NUREG and any
corrections are not substitutes for NRC
regulations, and compliance with them
is not required. The NUREG and
corrections describe approaches that are
generally acceptable to NRC staff.
However, methods and solutions
different than those in the NUREG and
corrections will be acceptable, if they
provide a basis for concluding that the
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decommissioning actions are in
compliance with NRC regulations.
Dated at Rockville, MD, this 10th day of
August, 2007.
For the Nuclear Regulatory Commission.
Keith I. McConnell,
Deputy Director, Decommissioning &
Uranium Recovery, Licensing Directorate,
Division of Waste Management and
Environmental Protection, Office of Federal
and State Materials and Environmental
Management Programs.
[FR Doc. E7–16131 Filed 8–15–07; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
Notice of Opportunity To Comment on
Model Safety Evaluation on Technical
Specification Improvement To Revise
Control Rod Notch Surveillance
Frequency, Clarify SRM Insert Control
Rod Action, and Clarify Frequency
Example
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 revision of Standard Technical
Specifications (STS), NUREG–1433
(BWR/4) and NUREG–1434 (BWR/6).
Specifically the SE addresses: (1) The
revision of the TS surveillance
requirement (SR) 3.1.3.2 frequency in
STS 3.1.3, ‘‘Control Rod
OPERABILITY,’’ (2) a clarification to the
requirement to fully insert all insertable
control rods for the limiting condition
for operation (LCO) in STS 3.3.1.2,
Required Action E.2, ‘‘Source Range
Monitor Instrumentation’’ (NUREG–
1434 only), and (3) the revision of
Example 1.4–3 in STS Section 1.4
‘‘Frequency’’ to clarify the applicability
of the 1.25 surveillance test interval
extension. The NRC staff has also
prepared a model license amendment
request and a model no significant
hazards consideration (NSHC)
determination relating to this matter.
The purpose of these models are to
permit the NRC to efficiently process
amendments that propose to modify TS
control rod SR testing frequency, clarify
TS control insertion requirements, and
clarify SR frequency discussions.
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 plant licensing
basis. The NRC staff is requesting
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comment on the model SE, model
amendment request, and model NSHC
determination prior to announcing their
availability for referencing in license
amendment applications.
DATES: The comment period expires
September 17, 2007. 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, Rulemaking, Directives, and
Editing 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
CLIIP@nrc.gov.
FOR FURTHER INFORMATION CONTACT:
Timothy Kobetz, Mail Stop: O–12H2,
Technical Specifications Branch,
Division of Inspection & 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 STS in a
manner that supports subsequent
license amendment applications. The
CLIIP includes an opportunity for the
public to comment on proposed changes
to the STS after a preliminary
assessment by the NRC staff and finding
that the change will likely be offered for
adoption by licensees. This notice
solicits comment on a proposed change
to the STS that modifies a TS control
rod SR testing frequency, clarifies TS
control rod insertion requirements, and
clarifies SR frequency discussions. 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
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availability of the change for adoption
by licensees. 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.
This notice involves the modification
of TS control rod SR testing frequency,
clarification of TS control insertion
requirements, and clarification of SR
frequency discussions. 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–475
Revision 1. TSTF–475 Revision 1 can be
viewed on the NRC’s Web page at
https://www.nrc.gov/reactors/operating/
licensing/techspecs.html.
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Applicability
This proposed TS change to modify
TS control rod SR testing frequency,
clarify TS control insertion
requirements, and clarify SR frequency
discussions is applicable to BWR NSSS
plants. The CLIIP does not prevent
licensees from requesting an alternative
approach or proposing the changes
without the attached model SE and the
NSHC. Variations from the approach
recommended in this notice may,
however, require additional review by
the NRC staff and may increase the time
and resources needed for by the NRC
staff and may increase the time and
resources needed for the review.
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, model
application 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
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hazards consideration determination,
and a notice of opportunity for a
hearing. The staff will also publish a
notice of issuance of an amendment to
operating license to announce the
modification of the TS control rod SR
testing frequency, TS control rod
insertion requirements, and SR
frequency discussions for each plant
that receives the requested change.
Proposed Safety Evaluation
Nuclear Regulatory Commission, Office
of Nuclear Reactor Regulation,
Consolidated Line Item Improvement
Program, Technical Specification Task,
Force (TSTF) Change TSTF–475,
Revision 1, Control Rod Notch Testing
Frequency and Source Range Monitor
Technical Specification Action To Insert
Control Rods
1.0 Introduction
By letter dated August 30, 2004, BWR
OWNERS Group (BWROG) submitted a
request for changes to NUREG–1433,
Standard Technical Specifications
General Electric Plants, BWR/4
(Reference 1), and NUREG–1434,
Standard Technical Specifications
General Electric Plants, BWR/6
(Reference 2). The proposed changes
would: (1) Revise the TS control rod
notch surveillance frequency in TS
3.1.3, ‘‘Control Rod OPERABILITY,’’ (2)
clarify the TS requirement for inserting
control rods for one or more inoperable
SRMs in MODE 5, and (3) revise one
Example in Section 1.4 ‘‘Frequency’’ to
clarify the applicability of the 1.25
surveillance test interval extension.
These changes are based on Technical
Specifications Task Force (TSTF)
change traveler TSTF–475, Revision 1,
that proposes revisions to the reference
BWR standard technical specifications
(STS) by: (1) Revising the frequency of
SR 3.1.3.2, notch testing of each fully
withdrawn control rod, from ‘‘7 days
after the control rod is withdrawn and
THERMAL POWER is greater than the
LPSP of RWM’’ to ‘‘31 days after the
control rod is withdrawn and
THERMAL POWER is greater than the
LPSP of the RWM’’, (2) adding the word
‘‘fully’’ to LCO 3.3.1.2 Required Action
E.2 (NUREG–1434 only) to clarify the
requirement to fully insert all insertable
control rods in core cells containing one
or more fuel assemblies when the
associated SRM instrument is
inoperable, and (3) revising Example
1.4–3 in Section 1.4 ‘‘Frequency’’ to
clarify that the 1.25 surveillance test
interval extension in SR 3.0.2 is
applicable to time periods discussed in
NOTES in the ‘‘SURVEILLANCE’’
column in addition to the time periods
in the ‘‘FREQUENCY’’ column.
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The purpose of these surveillances is
to confirm control rod insertion
capability which is demonstrated by
inserting each partially or fully
withdrawn control rod at least one
notch and observing that the control rod
moves. Control rods and control rod
drive (CRD) Mechanism (CRDM), by
which the control rods are moved, are
components of the CRD System, which
is the primary reactivity control system
for the reactor. By design, the CRDM is
highly reliable with a tapered design of
the index tube which is conducive to
control rod insertion.
A stuck control rod is an extremely
rare event and industry review of plant
operating experience did not identify
any incidents of stuck control rods
while performing a rod notch
surveillance test.
The purpose of these revisions is to
reduce the number of control rod
manipulations and, thereby, reduce the
opportunity for reactivity control
events.
2.0 Regulatory Evaluation
Title 10 of the Code of Federal
Regulations (CFR), Part 50, Appendix A,
General Design Criterion (GDC) 29,
Protection against anticipated
occurrence, requires that the protection
and reactivity control systems be
designed to assure an extremely high
probability of accomplishing their safety
functions in an event of anticipated
operational occurrences. The design
relies on the CRDS to function in
conjunction with the protection systems
under anticipated operational
occurrences, including loss of power to
all recirculation pumps, tripping of the
turbine generator, isolation of the main
condenser, and loss of all offsite power.
The CRDS provides an adequate means
of inserting sufficient negative reactivity
to shut down the reactor and prevent
exceeding acceptable fuel design limits
during anticipated operational
occurrences. Meeting the requirements
of GDC 29 for the CRDS prevents
occurrence of mechanisms that could
result in fuel cladding damage such as
severe overheating, excessive cladding
strain, or exceeding the thermal margin
limits during anticipated operational
occurrences. Preventing excessive
cladding damage in the event of
anticipated transients ensures
maintenance of the integrity of the
cladding as a fission product barrier.
3.0 Technical Evaluation
In order to perform this SE, the NRC
staff reviewed the following information
provided by the BWROG to justify the
submitted license amendment request
for STS NUREG–1433 and NUREG–1434
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to revise the weekly control rod notch
frequency to monthly, clarify the SRM
TS action for inserting control rods, and
the applicability of the 25% allowance
in Example 1.4–3. Specifically, the
following documents were reviewed
during the NRC staff’s evaluation:
• TSTF letter TSTF–04–07—Provided
a description of the proposed NUREG–
1433 and NUREG–1434 changes. TSTF–
475 would change the weekly rod notch
frequency to monthly, clarify the SRM
TS actions for inserting control rods,
and clarify the applicability of the 25%
allowance in Example 1.4–3 (Reference
3).
• TSTF letter TSTF–06–13—Provided
responses to NRC staff request for
additional information (RAI) on (1)
Industry experience with identifying
stuck rods, (2) tests that would identify
stuck rods, (3) continue compliance
with SIL 139, (4) industry experience on
collet failures, and (4) applying the 25%
grace period to the 31 day control rod
notch SR test frequency (Reference 4).
• BWROG letter BWROG–06036—
Provided the GE Nuclear Energy Report,
‘‘CRD Notching Surveillance Testing for
Limerick Generating Station,’’ in which
CRD notching frequency and CRD
performance were evaluated (Reference
5).
• TSTF letter TSTF–07–19—Provided
response to NRC staff RAI on CRD
performance in Control Cell Core (CCC)
designed plants, including TSTF–475,
Revision 1 (Reference 6).
The CRD System is the primary
reactivity control system for the reactor.
The CRD System, in conjunction with
the Reactor Protection System, provides
the means for the reliable control of
reactivity changes to ensure under all
conditions of normal operation,
including anticipated operational
occurrences that specified acceptable
fuel design limits are not exceeded.
Control rods are components of the CRD
System that have the capability to hold
the reactor core subcritical under all
conditions and to limit the potential
amount and rate of reactivity increase
caused by a malfunction in the CRD
System.
The CRD System consists of a CRDM,
by which the control rods are moved,
and a hydraulic control unit (HCU) for
each control rod. The CRDM is a
mechanical hydraulic latching cylinder
that positions the control blades. The
CRDM is a highly reliable mechanism
for inserting a control rod to the full-in
position. The collet piston mechanism
design feature ensures that the control
rod will not be inadvertently
withdrawn. This is accomplished by
engaging the collet fingers, mounted on
the collet piston, in notches located on
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the index tube. Due to the tapered
design of the index tube notches, the
collet piston mechanism will not
impede rod insertion under normal
insertion or scram conditions.
The collet retainer tube (CRT) is a
short tube welded to the upper end of
the CRD which houses the collet
mechanism which consist of the locking
collet, collet piston, collet return spring
and an unlocking cam. The collet
mechanism provides the locking/
unlocking mechanism that allows the
insert/withdraw movement of the
control rod. The CRT has three primary
functions: a) to carry the hydraulic
unlocking pressure to the collet piston,
b) to provide an outer cylinder, with a
suitable wear surface for the metal collet
piston rings, and c) to provide
mechanical support for the guide cap, a
component which incorporates the cam
surface for holding the collet fingers
open and also provides the upper rod
guide or bushing.
According to the BWROG, at the time
of the first CRT crack discovery in 1975
each partially or fully withdrawn
operable control rod was required to be
exercised one notch at least once each
week. It was recognized that notch
testing provided a method to
demonstrate the integrity of the CRT.
Control rod insertion capability was
demonstrated by inserting each partially
or fully withdrawn control rod at least
one notch and observing that the control
rod moves. The control rod may then be
returned to its original position. This
ensures the control rod is not stuck and
is free to insert on a scram signal.
It was determined that during scrams,
the CRT temperature distribution
changes substantially at reactor
operating conditions. Relatively cold
water moves upward through the inside
of the CRT and exits via the flow holes
into the annulus on the outside. At the
same time hot water from the reactor
vessel flows downward on the outside
surface of the CRT. There is very little
mixing of the cold water flowing from
the three flow holes into the annulus
and the hot water flowing downward.
Thus, there are substantial through wall
and circumferential temperature
gradients during scrams which
contribute to the observed CRT
cracking.
Subsequently, many BWRs have
reduced the frequency of notch testing
for partially withdrawn control rods
from weekly to monthly. The notch test
frequency for fully withdrawn control
rods are still performed weekly. The
change, for partially withdrawn control
rods, was made because of the potential
power reduction required to allow
control rod movement for partially
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withdrawn control rods, the desire to
coordinate scheduling with other plant
activities, and the fact that a large
sample of control rods are still notch
tested on the weekly basis. The
operating experience related to the
changes in CRD performance also
provided additional justification to
reduce the notch test frequency for the
partially withdrawn control rods.
In response to the NRC staff RAIs and
to support their position to reduce the
CRD notch testing frequency, the
BWROG provided plant data and GE
Nuclear Energy report, CRD Notching
Surveillance Testing for Limerick
Generating Station (CRDNST). The GE
report provided a description of the
cracks noted on the original design CRT
surfaces. These cracks, which were later
determined to be intergranular, were
generally circumferential, and appeared
with greatest frequency below and
between the cooling water ports, in the
area of the change in wall thickness.
Subsequently, cracks associated with
residual stresses were also observed in
the vicinity of the attachment weld.
Continued circumferential cracking
could lead to 360 degree severance of
the CRT that would render the CRD
inoperable which would prevent
insertion, withdrawal or scram. Such
failure would be detectable in any fully
or partially withdrawn control rod
during the surveillance notch testing
required by the Technical
Specifications. To a lesser degree, cracks
have also been noted at the welded joint
of the interim design CRT but no cracks
haven been observed in the final
improved CRT design. In a request for
additional information, BWROG
response of being unable to find a collet
housing failure since 1975 supported
the NRC staff review of not finding a
collet housing failure. To date, operating
experience data shows no reports of a
severed CRT at any BWR. No collet
housing failures have been noted since
1975. On a numerical basis for instance,
based on BWROG assumption that there
are 137 control rods for a typical BWR/
4 and 193 control rods for a typical
BWR/6, the yearly performance would
be 6590 rod notch tests for a BWR/4
plant and 9284 for a BWR/6 plant. For
example, if all BWRs operating in the
U.S. are taken into consideration, the
yearly performances of rod notch data
would translate into approximately
240,000 rod notch tests without
detecting a failure.
In addition, the IGSCC crack growth
rates were evaluated, at Limerick
Generating Station, using GE’s PLEDGE
model with the assumption that the
water chemistry condition is based on
GE recommendations. The model is
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based on fundamental principles of
stress corrosion cracking which can
evaluate crack growth rates as a function
of water oxygen level, conductivity,
material sensitization and applied loads.
It was determined that the additional
time of 24 days represented an
additional 10 mils of growth in total
crack length. The small difference in
growth rate would have little effect on
the behavior between one notch test and
the next subsequent test. Therefore,
from the materials perspective based on
low crack growth rates, a decrease in the
notch test frequency would not affect
the reliability of detecting a CRDM
failure due to crack growth.
Also, the BWR scram system has
extremely high reliability. In addition to
notch testing, scram time testing can
identify failure of individual CRD
operation resulting from IGSCC-initiated
cracks and mechanical binding. Unlike
the CRD notch tests, these single rod
scram tests cover the other mechanical
components such as scram pilot
solenoid operated valves, the scram
inlet and outlet air operated valves, and
the scram accumulator, as well as
operation of the control rods. Thus, the
primary assurance of scram system
reliability is provided by the scram time
testing since it monitors the system
scram operation and the complete travel
of the control rod.
Also, the HCUs, CRD drives, and
control rods are also tested during
refueling outages, approximately every
18–24 months. Based on the data
collected during the preceding cycle of
operation, selected control rod drives,
are inspected and, as required, their
internal components are replaced.
Therefore, increasing the CRD notch
testing frequency to monthly would
have very minimal impact on the
reliability of the scram system.
The NRC staff has reviewed the
BWROG TSTF’s proposal to amend the
TS SR 3.1.3.2, ‘‘Control Rod
OPERABILITY’’ from seven days to
monthly. Based on the following
evaluation condition: (1) Slow crack
growth rate of the CRT; (2) the improved
CRT design; (3) a higher reliable method
(scram time testing) to monitor CRD
scram system functionality; (4) GE
chemistry recommendations; and (5) no
known CRD failures have been detected
during the notch testing exercise, the
NRC staff concluded that the changes
would reduce the number of control rod
manipulations thereby reducing the
opportunity for potential reactivity
events while having a very minimal
impact on the extremely high reliability
of the CRD system. Therefore, the NRC
staff finds the change acceptable with
the commitment to implement GE water
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quality for the CRD system
recommendations. Furthermore, the
utilities should consider the
replacement of the CRT, when possible,
with the GE CRT improved design.
The NRC staff has reviewed the
BWROG TSTF–475 proposal to amend
the NUREG–1434, Specification 3.3.1.2,
Required Action E.2 from ‘‘Initiate
action to insert all insertable control
rods in core cells containing one or
more fuel assemblies’’ to ‘‘Initiate action
to fully insert all insertable control rods
in core cells containing one or more fuel
assemblies.’’ The NRC staff finds the
revision acceptable because the
requirement to insert control rods is
meant to require control rods to be fully
inserted and adding ‘‘fully’’ does not
change but clarifies the intent of the
action.
The NRC staff has reviewed the
BWROG TSTF–475 proposal to amend
Example 1.4–3 in Section 1.4
‘‘Frequency,’’ to make the 1.25
provision in SR 3.0.2 to be equally
applicable to time periods specified in
the ‘‘FREQUENCY’’ column and in the
NOTE in the ‘‘SURVEILLANCE’’
column. The NRC staff finds this change
acceptable since the revision would
make it consistent with the definition of
specified ‘‘Frequency’’ provided in the
second paragraph of Section 1.4 which
states that the specified ‘‘Frequency’’ is
referred to throughout this section and
each of the Specifications of Section 3.0,
Surveillance Requirement (SR)
Applicability. The specified
‘‘Frequency’’ consists of the
requirements of the Frequency column
of each SR, as well as certain Notes in
the Surveillance column that modify
performance requirements.
3.1 Conclusion
The NRC staff has reviewed the
licensee’s proposal to amend existing
TS sections SR 3.1.3.2, ‘‘Control Rod
OPERABILITY,’’ NUREG–1434, LCO
3.3.1.2 Required Action E.2, ‘‘Source
Range Monitor (SRM) Instrumentation,’’
and Example 1.4–3, ‘‘Frequency’’
applicable to SR 3.0.2. The NRC staff
has concluded that the TS revisions will
have a minimal affect on the high
reliability of the CRD system while
reducing the opportunity for potential
reactivity events; thus, meeting the
requirement of CFR, Part 50, Appendix
A, GDC 29. Therefore, the staff
concludes that the amendment request
is acceptable.
Based on the considerations discussed
above, the Commission has concluded
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
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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.
4.0 State Consultation
In accordance with the Commission’s
regulations, the [ ] State official was
notified of the proposed issuance of the
amendment. The State official had [(1)
No comments or (2) the following
comments—with subsequent
disposition by the staff].
5.0 Environmental Consideration
The amendments change a
requirement with respect to the
installation or use of a facility
component located within the restricted
area as defined in 10 CFR Part 20 and
change surveillance requirements. The
NRC staff has determined that the
amendments involve 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. The
Commission has previously issued a
proposed finding that the amendments
involve no significant hazards
considerations, and there has been no
public comment on the finding [FR ].
Accordingly, the amendments meet the
eligibility criteria for categorical
exclusion set forth in 10 CFR 51.22(c)(9)
[and (c)(10)]. Pursuant to 10 CFR
51.22(b), no environmental impact
statement or environmental assessment
need be prepared in connection with the
issuance of the amendments.
6.0 Conclusion
The Commission has concluded, on
the basis of 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, ‘‘Standard Technical
Specifications General Electric Plants, BWR/
4, Revision 3,’’ August 31, 2003.
2. NUREG–1434, ‘‘Standard Technical
Specifications General Electric Plants, BWR/
6, Revision 3,’’ August 31, 2003.
3. Letter TSTF–04–07 from the Technical
Specifications Task Force to the NRC, TSTF–
475 Revision 0, ‘‘Control Rod Notch Testing
Frequency and SRM Insert Control Rod
E:\FR\FM\16AUN1.SGM
16AUN1
Federal Register / Vol. 72, No. 158 / Thursday, August 16, 2007 / Notices
Action,’’ May 5, 2005, ADAMS accession
number ML042520035.
4. Letter TSTF–06–13 from the Technical
Specifications Task Force to the NRC,
‘‘Response to NRC Request for Additional
Information Regarding TSTF–475, Revision
0,’’ dated July 3, 2006, ADAMS accession
number ML0618403421.
5. Letter BWROG–06036 from the BWR
Owners Group to the NRC, ‘‘Response to NRC
Request for Additional Information
Regarding TSTF–475, Revision 0,’’ dated
November 16, 2006, Enclosure of the GE
Nuclear Energy Report, ‘‘CRD Notching
Surveillance Testing for Limerick Generating
Station,’’ dated November 2006, ADAMS
accession number ML0632502580.
6. Letter TSTF–07–19 from the Technical
Specifications Task Force to the NRC,
‘‘Response to NRC Request for Additional
Information Regarding TSTF–475, Revision
0,’’ dated May 22, 2007, ADAMS accession
number ML0714204280.
rwilkins on PROD1PC63 with NOTICES
The following example of an application
was prepared by the NRC staff to facilitate
use of the consolidated line item
improvement process (CLIIP). The model
provides the expected level of detail and
content for an application to revise technical
specifications regarding revision of control
rod notch surveillance test frequency,
clarification of SRM insert control rod action,
and a clarification of a frequency example.
Licensees remain responsible for ensuring
that their actual application fulfills their
administrative requirements as well as
Nuclear Regulatory Commission regulations.
U.S. Nuclear Regulatory Commission,
Document Control Desk,
Washington, DC 20555.
Subject:
PLANT NAME
DOCKET NO. 50–
APPLICATION FOR TECHNICAL
SPECIFICATION CHANGE REGARDING
REVISION OF CONTROL ROD NOTCH
SURVEILLANCE TEST FREQUENCY,
CLARIFICATION OF SRM INSERT
CONTROL ROD ACTION, AND A
CLARIFICATION OF A FREQUENCY
EXAMPLE USING THE CONSOLIDATED
LINE ITEM IMPROVEMENT PROCESS.
Gentleman: In accordance with the
provisions of 10 CFR 50.90 [LICENSEE] is
submitting a request for an amendment to the
technical specifications (TS) for [PLANT
NAME, UNIT NOS.].
The proposed amendment would: (1)
Revise the TS surveillance requirement (SR)
frequency in TS 3.1.3, ‘‘Control Rod
OPERABILITY’’, (2) clarify the requirement
to fully insert all insertable control rods for
the limiting condition for operation (LCO) in
TS 3.3.1.2, required Action E.2, ‘‘Source
Range Monitoring Instrumentation,’’ and (3)
revise Example 1.4–3 in Section 1.4
‘‘Frequency’’ to clarify the applicability of
the 1.25 surveillance test interval extension.
Attachment 1 provides a description of the
proposed change, the requested confirmation
of applicability, and plant-specific
verifications. Attachment 2 provides the
existing TS pages marked up to show the
proposed change. Attachment 3 provides
revised (clean) TS pages. Attachment 4
VerDate Aug<31>2005
17:27 Aug 15, 2007
Jkt 211001
provides a summary of the regulatory
commitments made in this submittal.
[LICENSEE] requests approval of the
proposed License Amendment by [DATE],
with the amendment being implemented [BY
DATE OR WITHIN X DAYS].
In accordance with 10 CFR 50.91, a copy
of this application, with attachments, is being
provided to the designated [STATE] Official.
I declare under penalty of perjury under
the laws of the United Stats of America that
I am authorized by [LICENSEE] to make this
request and that the foregoing is true and
correct. (Note that request may be notarized
in lieu of using this oath or affirmation
statement). If you should have any questions
regarding this submittal, please contact
[NAME, TELEPHONE NUMBER]
Sincerely,
[Name, Title]
Attachments:
1. Description and Assessment.
2. Proposed Technical Specification
Changes.
3. Revised Technical Specification Pages.
4. Regulatory Commitments.
5. Proposed Technical Specification Bases
Changes.
CC: NRC Project Manager.
NRC Regional Office.
NRC Resident Inspector.
State Contact.
Attachment 1—Description and
Assessment
1.0 Description
The proposed amendment would: (1)
Revise the TS surveillance requirement
(SR 3.1.3.2) frequency in TS 3.1.3,
‘‘Control Rod OPERABILITY’’, (2)
clarify the requirement to fully insert all
insertable control rods for the limiting
condition for operation (LCO) in TS
3.3.1.2, Required Action E.2, ‘‘Source
Range Monitoring Instrumentation’’,
and (3) revise Example 1.4–3 in Section
1.4 ‘‘Frequency’’ to clarify the
applicability of the 1.25 surveillance
test interval extension.
The changes are consistent with
Nuclear Regulatory Commission (NRC)
approved Industry/Technical
Specification Task Force (TSTF) STS
change TSTF–475, Revision 1. The
Federal Register notice published on
[DATE] announced the availability of
this TS improvement through the
consolidated line item improvement
process (CLIIP).
2.0
Assessment
2.1 Applicability of Published Safety
Evaluation
[LICENSEE] has reviewed the safety
evaluation dated [DATE] as part of the
CLIIP. This review included a review of
the NRC staff’s evaluation, as well as the
supporting information provided to
support TSTF–475, Revision 1.
[LICENSEE] has concluded that the
justifications presented in the TSTF
PO 00000
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Fmt 4703
Sfmt 4703
46107
proposal and the safety evaluation
prepared by the NRC staff are applicable
to [PLANT, UNIT NOS.] and justify this
amendment for the incorporation of the
changes to the [PLANT] TS.
2.2 Optional Changes and Variations
[LICENSEE] is not proposing any
variations or deviations from the TS
changes described in the modified
TSTF–475, Revision 1 and the NRC
staff’s model safety evaluation dated
[DATE].
3.0
Regulatory Analysis
3.1 No Significant Hazards
Consideration Determination
[LICENSEE] has reviewed the
proposed no significant hazards
consideration determination (NSHCD)
published in the Federal Register as
part of the CLIIP. [LICENSEE] has
concluded that the proposed NSHCD
presented in the Federal Register notice
is applicable to [PLANT] and is hereby
incorporated by reference to satisfy the
requirements of 10 CFR 50.91(a).
3.2 Verification and Commitments
As discussed in the notice of
availability published in the Federal
Register on [DATE] for this TS
improvement, the [LICENSEE] verifies
the applicability of TSTF–475 to
[PLANT], and commits to establishing
Technical Specification Bases for TS as
proposed in TSTF–475, Revision 1.
These changes are based on TSTF
change traveler TSTF–475 (Revision 1)
that proposes revisions to the BWR STS
by: (1) Revising the frequency of SR
3.1.3.2, notch testing of fully withdrawn
control rod, from ‘‘7 days after the
control rod is withdrawn and
THERMAL POWER is greater than the
LPSP of RWM’’ to ‘‘31 days after the
control rod is withdrawn and
THERMAL POWER is greater than the
LPSP of the RWM’’, (2) adding the word
‘‘fully’’ to LCO 3.3.1.2 Required Action
E.2 (NUREG–1434 only) to clarify the
requirement to fully insert all insertable
control rods in core cells containing one
or more fuel assemblies when the
associated SRM instrument is
inoperable, and (3) revising Example
1.4–3 in Section 1.4 ‘‘Frequency’’ to
clarify that the 1.25 surveillance test
interval extension in SR 3.0.2 is
applicable to time periods discussed in
NOTES in the ‘‘SURVEILLANCE’’
column in addition to the time periods
in the ‘‘FREQUENCY’’ column.
4.0 Environmental Evaluation
[LICENSEE] has reviewed the
environmental evaluation included in
the model safety evaluation dated
[DATE] as part of the CLIIP. [LICENSEE]
E:\FR\FM\16AUN1.SGM
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Federal Register / Vol. 72, No. 158 / Thursday, August 16, 2007 / Notices
has concluded that the staff’s findings
presented in that evaluation are
applicable to [PLANT] and the
evaluation is hereby incorporated by
reference for this application.
Attachment 2—Proposed Technical
Specification Changes (Mark-Up)
Attachment 3—Proposed Technical
Specification Pages
Attachment 4—List of Regulatory
Commitments
this document. Any other statements in
this submittal are provided for
information purposes and are not
considered to be regulatory
commitments. Please direct questions
regarding these commitments to
[CONTACT NAME].
The following table identifies those
actions committed to by [LICENSEE] in
Regulatory commitments
Due date/event
[LICENSEE] will establish the Technical Specification Bases for [TS B 3.1.3, TS B 3.1.4, and
TS B 3.3.1.2] as adopted with the applicable license amendment.
[LICENSEE] will establish the water quality controls as recommended by SIL No. 148, Water
Quality Control for the Control Rod System,’’ September 15, 1975.
hazards consideration is presented
below:
Proposed No Significant Hazards
Consideration Determination
rwilkins on PROD1PC63 with NOTICES
Attachment 5—Proposed Changes to
Technical Specification Bases Pages
Criterion 1—The Proposed Change Does
Not Involve a Significant Increase in the
Probability or Consequences of an
Accident Previously Evaluated
The proposed change generically
implements TSTF–475, Revision 1,
‘‘Control Rod Notch Testing Frequency
and SRM Insert Control Rod Action.’’
TSTF–475, Revision 1 modifies
NUREG–1433 (BWR/4) and NUREG–
1434 (BWR/6) STS. The changes: (1)
Revise TS testing frequency for
surveillance requirement (SR) 3.1.3.2 in
TS 3.1.3, ‘‘Control Rod OPERABILITY’’,
(2) clarify the requirement to fully insert
all insertable control rods for the
limiting condition for operation (LCO)
in TS 3.3.1.2, Required Action E.2,
‘‘Source Range Monitoring
Instrumentation’’ (NUREG–1434 only),
and (3) revise Example 1.4–3 in Section
1.4 ‘‘Frequency’’ to clarify the
applicability of the 1.25 surveillance
test interval extension. The
consequences of an accident after
adopting TSTF–475, Revision 1 are no
different than the consequences of an
accident prior to adoption. Therefore,
this change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
Description of Amendment Request:
[Plant Name] requests adoption of an
approved change to the Standard
Technical Specifications (STS) for
General Electric (GE) Plants (NUREG–
1433, BWR/4 and NUREG–1434, BWR/
6) and plant specific technical
specifications (TS), that allows: (1)
Revising the frequency of SR 3.1.3.2,
notch testing of fully withdrawn control
rod, from ‘‘7 days after the control rod
is withdrawn and THERMAL POWER is
greater than the LPSP of RWM’’ to ‘‘31
days after the control rod is withdrawn
and THERMAL POWER is greater than
the LPSP of the RWM’’, (2) adding the
word ‘‘fully’’ to LCO 3.3.1.2 Required
Action E.2 (NUREG–1434 only) to
clarify the requirement to fully insert all
insertable control rods in core cells
containing one or more fuel assemblies
when the associated SRM instrument is
inoperable, and (3) revising Example
1.4–3 in Section 1.4 ‘‘Frequency’’ to
clarify that the 1.25 surveillance test
interval extension in SR 3.0.2 is
applicable to time periods discussed in
NOTES in the ‘‘SURVEILLANCE’’
column in addition to the time periods
in the ‘‘FREQUENCY’’ column. The staff
finds that the proposed STS changes are
acceptable because the number of
control rod manipulations is reduced
thereby reducing the opportunity for
potential reactivity events while having
a very minimal impact on the extremely
high reliability of the CRD system as
discussed in the technical evaluation
section of this safety evaluation.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), an
analysis of the issue of no significant
VerDate Aug<31>2005
17:27 Aug 15, 2007
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Criterion 2—The Proposed Change Does
Not Create the Possibility of a New or
Different Kind of Accident From any
Previously Evaluated
The proposed change does not
involve a physical alteration of the plant
(no new or different type of equipment
will be installed) or a change in the
methods governing normal plant
operation. The proposed change will not
introduce new failure modes or effects
and will not, in the absence of other
unrelated failures, lead to an accident
PO 00000
Frm 00077
Fmt 4703
Sfmt 4703
[Complete,
within X
ment].
[Complete,
within X
ment].
implemented with amendment or
days of implementation of amendimplemented with amendment or
days of implementation of amend-
whose consequences exceed the
consequences of accidents previously
analyzed. Thus, this change does not
create the possibility of a new or
different kind of accident from any
accident previously evaluated.
Criterion 3—The Proposed Change Does
Not Involve a Significant Reduction in
the Margin of Safety
TSTF–475, Revision 1 will: (1) Revise
the TS SR 3.1.3.2 frequency in TS 3.1.3,
‘‘Control Rod OPERABILITY’’, (2)
clarify the requirement to fully insert all
insertable control rods for the limiting
condition for operation (LCO) in TS
3.3.1.2, ‘‘Source Range Monitoring
Instrumentation’’ (NUREG–1434 only),
and (3) revise Example 1.4–3 in Section
1.4 ‘‘Frequency’’ to clarify the
applicability of the 1.25 surveillance
test interval extension. The GE Nuclear
Energy Report, ‘‘CRD Notching
Surveillance Testing for Limerick
Generating Station,’’ dated November
2006, concludes that extending the
control rod notch test interval from
weekly to monthly is not expected to
impact the reliability of the scram
system and that the analysis supports
the decision to change the surveillance
frequency. Therefore, the proposed
changes in TSTF–475, Revision 1 are
acceptable and do not involve a
significant reduction in a margin of
safety.
Based upon the reasoning presented
above and the previous discussion of
the amendment request, the requested
change does not involve a significant
hazards consideration.
Dated at Rockville, Maryland, this 9th day
of August, 2007.
E:\FR\FM\16AUN1.SGM
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Federal Register / Vol. 72, No. 158 / Thursday, August 16, 2007 / Notices
For the Nuclear Regulatory Commission.
Carl Schulten,
Acting Chief, Technical Specifications
Branch, Division of Inspection & Regional
Support, Office of Nuclear Reactor
Regulation.
[FR Doc. E7–16138 Filed 8–15–07; 8:45 am]
BILLING CODE 7590–01–P
PENSION BENEFIT GUARANTY
CORPORATION
Submission of Information Collection
for OMB Review; Comment Request;
Filings for Reconsiderations
Pension Benefit Guaranty
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ACTION: Notice of request for OMB
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AGENCY:
SUMMARY: Pension Benefit Guaranty
Corporation (PBGC) is requesting that
the Office of Management and Budget
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Rules for Administrative Review of
Agency Decisions. This notice informs
the public of PBGC’s request and solicits
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information.
Comments should be submitted
by September 17, 2007.
ADDRESSES: Comments should be sent to
the Office of Information and Regulatory
Affairs, Office of Management and
Budget, Attention: Desk Officer for
Pension Benefit Guaranty Corporation,
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OIRA_DOCKET@omb.eop.gov or by fax
to (202) 395–6974. Copies of the
collection of information may also be
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(TTY and TDD users may call the
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DATES:
rwilkins on PROD1PC63 with NOTICES
FOR FURTHER INFORMATION CONTACT:
Donald F. McCabe, Attorney, Legislative
and Regulatory Department, Pension
Benefit Guaranty Corporation, 1200 K
Street, NW., Washington, DC 20005–
4026, 202–326–4024. (For TTY/TDD
users, call the Federal relay service tollfree at 1–800–877–8339 and ask to be
connected to 202–326–4024.)
VerDate Aug<31>2005
17:27 Aug 15, 2007
Jkt 211001
PBGC’s
regulation on Rules for Administrative
Review of Agency Decisions (29 CFR
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specifies which initial determinations
are subject to reconsideration. Subpart C
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request, where to submit it, form and
content of reconsideration requests, and
other matters relating to
reconsiderations.
Any person aggrieved by an initial
determination of PBGC under
4003.1(b)(1) (determinations that a plan
is covered by section 4021 of ERISA),
4003.1(b)(2) (determinations concerning
premiums, interest, and late payment
penalties under section 4007 of ERISA),
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§ 4003.1(b)(4) (determinations
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SUPPLEMENTARY INFORMATION:
Issued in Washington, DC, this 10th day of
August, 2007.
John H. Hanley,
Director, Legislative and Regulatory
Department, Pension Benefit Guaranty
Corporation.
[FR Doc. E7–16101 Filed 8–15–07; 8:45 am]
BILLING CODE 7709–01–P
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46109
PENSION BENEFIT GUARANTY
CORPORATION
Agency Information Collection
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Pension Benefit Guaranty
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ACTION: Notice of request for extension
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AGENCY:
SUMMARY: The Pension Benefit Guaranty
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DATES: Comments should be submitted
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ADDRESSES: Comments should be sent to
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Affairs, Office of Management and
Budget, Attention: Desk Officer for
Pension Benefit Guaranty Corporation,
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Copies of the request for extension
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may be obtained without charge by
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at the above address, visiting the
Disclosure Division, faxing a request to
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E:\FR\FM\16AUN1.SGM
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]]>Agencies
[Federal Register Volume 72, Number 158 (Thursday, August 16, 2007)]
[Notices]
[Pages 46103-46109]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-16138]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Notice of Opportunity To Comment on Model Safety Evaluation on
Technical Specification Improvement To Revise Control Rod Notch
Surveillance Frequency, Clarify SRM Insert Control Rod Action, and
Clarify Frequency Example
AGENCY: Nuclear Regulatory Commission.
ACTION: Request for comment.
-----------------------------------------------------------------------
SUMMARY: Notice is hereby given that the staff of the Nuclear
Regulatory Commission (NRC) has prepared a model safety evaluation (SE)
relating to the revision of Standard Technical Specifications (STS),
NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6). Specifically the SE
addresses: (1) The revision of the TS surveillance requirement (SR)
3.1.3.2 frequency in STS 3.1.3, ``Control Rod OPERABILITY,'' (2) a
clarification to the requirement to fully insert all insertable control
rods for the limiting condition for operation (LCO) in STS 3.3.1.2,
Required Action E.2, ``Source Range Monitor Instrumentation'' (NUREG-
1434 only), and (3) the revision of Example 1.4-3 in STS Section 1.4
``Frequency'' to clarify the applicability of the 1.25 surveillance
test interval extension. The NRC staff has also prepared a model
license amendment request and a model no significant hazards
consideration (NSHC) determination relating to this matter. The purpose
of these models are to permit the NRC to efficiently process amendments
that propose to modify TS control rod SR testing frequency, clarify TS
control insertion requirements, and clarify SR frequency discussions.
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 plant licensing basis. The NRC staff is
requesting comment on the model SE, model amendment request, and model
NSHC determination prior to announcing their availability for
referencing in license amendment applications.
DATES: The comment period expires September 17, 2007. 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, Rulemaking, Directives, and
Editing 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
CLIIP@nrc.gov.
FOR FURTHER INFORMATION CONTACT: Timothy Kobetz, Mail Stop: O-12H2,
Technical Specifications Branch, Division of Inspection & 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 STS in a manner that supports subsequent
license amendment applications. The CLIIP includes an opportunity for
the public to comment on proposed changes to the STS after a
preliminary assessment by the NRC staff and finding that the change
will likely be offered for adoption by licensees. This notice solicits
comment on a proposed change to the STS that modifies a TS control rod
SR testing frequency, clarifies TS control rod insertion requirements,
and clarifies SR frequency discussions. 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
[[Page 46104]]
availability of the change for adoption by licensees. 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.
This notice involves the modification of TS control rod SR testing
frequency, clarification of TS control insertion requirements, and
clarification of SR frequency discussions. 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-475 Revision 1. TSTF-475 Revision 1 can be
viewed on the NRC's Web page at https://www.nrc.gov/reactors/operating/
licensing/techspecs.html.
Applicability
This proposed TS change to modify TS control rod SR testing
frequency, clarify TS control insertion requirements, and clarify SR
frequency discussions is applicable to BWR NSSS plants. The CLIIP does
not prevent licensees from requesting an alternative approach or
proposing the changes without the attached model SE and the NSHC.
Variations from the approach recommended in this notice may, however,
require additional review by the NRC staff and may increase the time
and resources needed for by the NRC staff and may increase the time and
resources needed for the review.
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, model application 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 operating license to announce the
modification of the TS control rod SR testing frequency, TS control rod
insertion requirements, and SR frequency discussions for each plant
that receives the requested change.
Proposed Safety Evaluation
Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,
Consolidated Line Item Improvement Program, Technical Specification
Task, Force (TSTF) Change TSTF-475, Revision 1, Control Rod Notch
Testing Frequency and Source Range Monitor Technical Specification
Action To Insert Control Rods
1.0 Introduction
By letter dated August 30, 2004, BWR OWNERS Group (BWROG) submitted
a request for changes to NUREG-1433, Standard Technical Specifications
General Electric Plants, BWR/4 (Reference 1), and NUREG-1434, Standard
Technical Specifications General Electric Plants, BWR/6 (Reference 2).
The proposed changes would: (1) Revise the TS control rod notch
surveillance frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2)
clarify the TS requirement for inserting control rods for one or more
inoperable SRMs in MODE 5, and (3) revise one Example in Section 1.4
``Frequency'' to clarify the applicability of the 1.25 surveillance
test interval extension.
These changes are based on Technical Specifications Task Force
(TSTF) change traveler TSTF-475, Revision 1, that proposes revisions to
the reference BWR standard technical specifications (STS) by: (1)
Revising the frequency of SR 3.1.3.2, notch testing of each fully
withdrawn control rod, from ``7 days after the control rod is withdrawn
and THERMAL POWER is greater than the LPSP of RWM'' to ``31 days after
the control rod is withdrawn and THERMAL POWER is greater than the LPSP
of the RWM'', (2) adding the word ``fully'' to LCO 3.3.1.2 Required
Action E.2 (NUREG-1434 only) to clarify the requirement to fully insert
all insertable control rods in core cells containing one or more fuel
assemblies when the associated SRM instrument is inoperable, and (3)
revising Example 1.4-3 in Section 1.4 ``Frequency'' to clarify that the
1.25 surveillance test interval extension in SR 3.0.2 is applicable to
time periods discussed in NOTES in the ``SURVEILLANCE'' column in
addition to the time periods in the ``FREQUENCY'' column.
The purpose of these surveillances is to confirm control rod
insertion capability which is demonstrated by inserting each partially
or fully withdrawn control rod at least one notch and observing that
the control rod moves. Control rods and control rod drive (CRD)
Mechanism (CRDM), by which the control rods are moved, are components
of the CRD System, which is the primary reactivity control system for
the reactor. By design, the CRDM is highly reliable with a tapered
design of the index tube which is conducive to control rod insertion.
A stuck control rod is an extremely rare event and industry review
of plant operating experience did not identify any incidents of stuck
control rods while performing a rod notch surveillance test.
The purpose of these revisions is to reduce the number of control
rod manipulations and, thereby, reduce the opportunity for reactivity
control events.
2.0 Regulatory Evaluation
Title 10 of the Code of Federal Regulations (CFR), Part 50,
Appendix A, General Design Criterion (GDC) 29, Protection against
anticipated occurrence, requires that the protection and reactivity
control systems be designed to assure an extremely high probability of
accomplishing their safety functions in an event of anticipated
operational occurrences. The design relies on the CRDS to function in
conjunction with the protection systems under anticipated operational
occurrences, including loss of power to all recirculation pumps,
tripping of the turbine generator, isolation of the main condenser, and
loss of all offsite power. The CRDS provides an adequate means of
inserting sufficient negative reactivity to shut down the reactor and
prevent exceeding acceptable fuel design limits during anticipated
operational occurrences. Meeting the requirements of GDC 29 for the
CRDS prevents occurrence of mechanisms that could result in fuel
cladding damage such as severe overheating, excessive cladding strain,
or exceeding the thermal margin limits during anticipated operational
occurrences. Preventing excessive cladding damage in the event of
anticipated transients ensures maintenance of the integrity of the
cladding as a fission product barrier.
3.0 Technical Evaluation
In order to perform this SE, the NRC staff reviewed the following
information provided by the BWROG to justify the submitted license
amendment request for STS NUREG-1433 and NUREG-1434
[[Page 46105]]
to revise the weekly control rod notch frequency to monthly, clarify
the SRM TS action for inserting control rods, and the applicability of
the 25% allowance in Example 1.4-3. Specifically, the following
documents were reviewed during the NRC staff's evaluation:
TSTF letter TSTF-04-07--Provided a description of the
proposed NUREG-1433 and NUREG-1434 changes. TSTF-475 would change the
weekly rod notch frequency to monthly, clarify the SRM TS actions for
inserting control rods, and clarify the applicability of the 25%
allowance in Example 1.4-3 (Reference 3).
TSTF letter TSTF-06-13--Provided responses to NRC staff
request for additional information (RAI) on (1) Industry experience
with identifying stuck rods, (2) tests that would identify stuck rods,
(3) continue compliance with SIL 139, (4) industry experience on collet
failures, and (4) applying the 25% grace period to the 31 day control
rod notch SR test frequency (Reference 4).
BWROG letter BWROG-06036--Provided the GE Nuclear Energy
Report, ``CRD Notching Surveillance Testing for Limerick Generating
Station,'' in which CRD notching frequency and CRD performance were
evaluated (Reference 5).
TSTF letter TSTF-07-19--Provided response to NRC staff RAI
on CRD performance in Control Cell Core (CCC) designed plants,
including TSTF-475, Revision 1 (Reference 6).
The CRD System is the primary reactivity control system for the
reactor. The CRD System, in conjunction with the Reactor Protection
System, provides the means for the reliable control of reactivity
changes to ensure under all conditions of normal operation, including
anticipated operational occurrences that specified acceptable fuel
design limits are not exceeded. Control rods are components of the CRD
System that have the capability to hold the reactor core subcritical
under all conditions and to limit the potential amount and rate of
reactivity increase caused by a malfunction in the CRD System.
The CRD System consists of a CRDM, by which the control rods are
moved, and a hydraulic control unit (HCU) for each control rod. The
CRDM is a mechanical hydraulic latching cylinder that positions the
control blades. The CRDM is a highly reliable mechanism for inserting a
control rod to the full-in position. The collet piston mechanism design
feature ensures that the control rod will not be inadvertently
withdrawn. This is accomplished by engaging the collet fingers, mounted
on the collet piston, in notches located on the index tube. Due to the
tapered design of the index tube notches, the collet piston mechanism
will not impede rod insertion under normal insertion or scram
conditions.
The collet retainer tube (CRT) is a short tube welded to the upper
end of the CRD which houses the collet mechanism which consist of the
locking collet, collet piston, collet return spring and an unlocking
cam. The collet mechanism provides the locking/unlocking mechanism that
allows the insert/withdraw movement of the control rod. The CRT has
three primary functions: a) to carry the hydraulic unlocking pressure
to the collet piston, b) to provide an outer cylinder, with a suitable
wear surface for the metal collet piston rings, and c) to provide
mechanical support for the guide cap, a component which incorporates
the cam surface for holding the collet fingers open and also provides
the upper rod guide or bushing.
According to the BWROG, at the time of the first CRT crack
discovery in 1975 each partially or fully withdrawn operable control
rod was required to be exercised one notch at least once each week. It
was recognized that notch testing provided a method to demonstrate the
integrity of the CRT. Control rod insertion capability was demonstrated
by inserting each partially or fully withdrawn control rod at least one
notch and observing that the control rod moves. The control rod may
then be returned to its original position. This ensures the control rod
is not stuck and is free to insert on a scram signal.
It was determined that during scrams, the CRT temperature
distribution changes substantially at reactor operating conditions.
Relatively cold water moves upward through the inside of the CRT and
exits via the flow holes into the annulus on the outside. At the same
time hot water from the reactor vessel flows downward on the outside
surface of the CRT. There is very little mixing of the cold water
flowing from the three flow holes into the annulus and the hot water
flowing downward. Thus, there are substantial through wall and
circumferential temperature gradients during scrams which contribute to
the observed CRT cracking.
Subsequently, many BWRs have reduced the frequency of notch testing
for partially withdrawn control rods from weekly to monthly. The notch
test frequency for fully withdrawn control rods are still performed
weekly. The change, for partially withdrawn control rods, was made
because of the potential power reduction required to allow control rod
movement for partially withdrawn control rods, the desire to coordinate
scheduling with other plant activities, and the fact that a large
sample of control rods are still notch tested on the weekly basis. The
operating experience related to the changes in CRD performance also
provided additional justification to reduce the notch test frequency
for the partially withdrawn control rods.
In response to the NRC staff RAIs and to support their position to
reduce the CRD notch testing frequency, the BWROG provided plant data
and GE Nuclear Energy report, CRD Notching Surveillance Testing for
Limerick Generating Station (CRDNST). The GE report provided a
description of the cracks noted on the original design CRT surfaces.
These cracks, which were later determined to be intergranular, were
generally circumferential, and appeared with greatest frequency below
and between the cooling water ports, in the area of the change in wall
thickness. Subsequently, cracks associated with residual stresses were
also observed in the vicinity of the attachment weld. Continued
circumferential cracking could lead to 360 degree severance of the CRT
that would render the CRD inoperable which would prevent insertion,
withdrawal or scram. Such failure would be detectable in any fully or
partially withdrawn control rod during the surveillance notch testing
required by the Technical Specifications. To a lesser degree, cracks
have also been noted at the welded joint of the interim design CRT but
no cracks haven been observed in the final improved CRT design. In a
request for additional information, BWROG response of being unable to
find a collet housing failure since 1975 supported the NRC staff review
of not finding a collet housing failure. To date, operating experience
data shows no reports of a severed CRT at any BWR. No collet housing
failures have been noted since 1975. On a numerical basis for instance,
based on BWROG assumption that there are 137 control rods for a typical
BWR/4 and 193 control rods for a typical BWR/6, the yearly performance
would be 6590 rod notch tests for a BWR/4 plant and 9284 for a BWR/6
plant. For example, if all BWRs operating in the U.S. are taken into
consideration, the yearly performances of rod notch data would
translate into approximately 240,000 rod notch tests without detecting
a failure.
In addition, the IGSCC crack growth rates were evaluated, at
Limerick Generating Station, using GE's PLEDGE model with the
assumption that the water chemistry condition is based on GE
recommendations. The model is
[[Page 46106]]
based on fundamental principles of stress corrosion cracking which can
evaluate crack growth rates as a function of water oxygen level,
conductivity, material sensitization and applied loads. It was
determined that the additional time of 24 days represented an
additional 10 mils of growth in total crack length. The small
difference in growth rate would have little effect on the behavior
between one notch test and the next subsequent test. Therefore, from
the materials perspective based on low crack growth rates, a decrease
in the notch test frequency would not affect the reliability of
detecting a CRDM failure due to crack growth.
Also, the BWR scram system has extremely high reliability. In
addition to notch testing, scram time testing can identify failure of
individual CRD operation resulting from IGSCC-initiated cracks and
mechanical binding. Unlike the CRD notch tests, these single rod scram
tests cover the other mechanical components such as scram pilot
solenoid operated valves, the scram inlet and outlet air operated
valves, and the scram accumulator, as well as operation of the control
rods. Thus, the primary assurance of scram system reliability is
provided by the scram time testing since it monitors the system scram
operation and the complete travel of the control rod.
Also, the HCUs, CRD drives, and control rods are also tested during
refueling outages, approximately every 18-24 months. Based on the data
collected during the preceding cycle of operation, selected control rod
drives, are inspected and, as required, their internal components are
replaced. Therefore, increasing the CRD notch testing frequency to
monthly would have very minimal impact on the reliability of the scram
system.
The NRC staff has reviewed the BWROG TSTF's proposal to amend the
TS SR 3.1.3.2, ``Control Rod OPERABILITY'' from seven days to monthly.
Based on the following evaluation condition: (1) Slow crack growth rate
of the CRT; (2) the improved CRT design; (3) a higher reliable method
(scram time testing) to monitor CRD scram system functionality; (4) GE
chemistry recommendations; and (5) no known CRD failures have been
detected during the notch testing exercise, the NRC staff concluded
that the changes would reduce the number of control rod manipulations
thereby reducing the opportunity for potential reactivity events while
having a very minimal impact on the extremely high reliability of the
CRD system. Therefore, the NRC staff finds the change acceptable with
the commitment to implement GE water quality for the CRD system
recommendations. Furthermore, the utilities should consider the
replacement of the CRT, when possible, with the GE CRT improved design.
The NRC staff has reviewed the BWROG TSTF-475 proposal to amend the
NUREG-1434, Specification 3.3.1.2, Required Action E.2 from ``Initiate
action to insert all insertable control rods in core cells containing
one or more fuel assemblies'' to ``Initiate action to fully insert all
insertable control rods in core cells containing one or more fuel
assemblies.'' The NRC staff finds the revision acceptable because the
requirement to insert control rods is meant to require control rods to
be fully inserted and adding ``fully'' does not change but clarifies
the intent of the action.
The NRC staff has reviewed the BWROG TSTF-475 proposal to amend
Example 1.4-3 in Section 1.4 ``Frequency,'' to make the 1.25 provision
in SR 3.0.2 to be equally applicable to time periods specified in the
``FREQUENCY'' column and in the NOTE in the ``SURVEILLANCE'' column.
The NRC staff finds this change acceptable since the revision would
make it consistent with the definition of specified ``Frequency''
provided in the second paragraph of Section 1.4 which states that the
specified ``Frequency'' is referred to throughout this section and each
of the Specifications of Section 3.0, Surveillance Requirement (SR)
Applicability. The specified ``Frequency'' consists of the requirements
of the Frequency column of each SR, as well as certain Notes in the
Surveillance column that modify performance requirements.
3.1 Conclusion
The NRC staff has reviewed the licensee's proposal to amend
existing TS sections SR 3.1.3.2, ``Control Rod OPERABILITY,'' NUREG-
1434, LCO 3.3.1.2 Required Action E.2, ``Source Range Monitor (SRM)
Instrumentation,'' and Example 1.4-3, ``Frequency'' applicable to SR
3.0.2. The NRC staff has concluded that the TS revisions will have a
minimal affect on the high reliability of the CRD system while reducing
the opportunity for potential reactivity events; thus, meeting the
requirement of CFR, Part 50, Appendix A, GDC 29. Therefore, the staff
concludes that the amendment request is acceptable.
Based on the considerations discussed above, the Commission has
concluded 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.
4.0 State Consultation
In accordance with the Commission's regulations, the [ ] State
official was notified of the proposed issuance of the amendment. The
State official had [(1) No comments or (2) the following comments--with
subsequent disposition by the staff].
5.0 Environmental Consideration
The amendments change a requirement with respect to the
installation or use of a facility component located within the
restricted area as defined in 10 CFR Part 20 and change surveillance
requirements. The NRC staff has determined that the amendments involve
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. The Commission has previously issued a proposed
finding that the amendments involve no significant hazards
considerations, and there has been no public comment on the finding [FR
]. Accordingly, the amendments meet the eligibility criteria for
categorical exclusion set forth in 10 CFR 51.22(c)(9) [and (c)(10)].
Pursuant to 10 CFR 51.22(b), no environmental impact statement or
environmental assessment need be prepared in connection with the
issuance of the amendments.
6.0 Conclusion
The Commission has concluded, on the basis of 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, ``Standard Technical Specifications General
Electric Plants, BWR/4, Revision 3,'' August 31, 2003.
2. NUREG-1434, ``Standard Technical Specifications General
Electric Plants, BWR/6, Revision 3,'' August 31, 2003.
3. Letter TSTF-04-07 from the Technical Specifications Task
Force to the NRC, TSTF-475 Revision 0, ``Control Rod Notch Testing
Frequency and SRM Insert Control Rod
[[Page 46107]]
Action,'' May 5, 2005, ADAMS accession number ML042520035.
4. Letter TSTF-06-13 from the Technical Specifications Task
Force to the NRC, ``Response to NRC Request for Additional
Information Regarding TSTF-475, Revision 0,'' dated July 3, 2006,
ADAMS accession number ML0618403421.
5. Letter BWROG-06036 from the BWR Owners Group to the NRC,
``Response to NRC Request for Additional Information Regarding TSTF-
475, Revision 0,'' dated November 16, 2006, Enclosure of the GE
Nuclear Energy Report, ``CRD Notching Surveillance Testing for
Limerick Generating Station,'' dated November 2006, ADAMS accession
number ML0632502580.
6. Letter TSTF-07-19 from the Technical Specifications Task
Force to the NRC, ``Response to NRC Request for Additional
Information Regarding TSTF-475, Revision 0,'' dated May 22, 2007,
ADAMS accession number ML0714204280.
The following example of an application was prepared by the NRC
staff to facilitate use of the consolidated line item improvement
process (CLIIP). The model provides the expected level of detail and
content for an application to revise technical specifications
regarding revision of control rod notch surveillance test frequency,
clarification of SRM insert control rod action, and a clarification
of a frequency example. Licensees remain responsible for ensuring
that their actual application fulfills their administrative
requirements as well as Nuclear Regulatory Commission regulations.
U.S. Nuclear Regulatory Commission,
Document Control Desk,
Washington, DC 20555.
Subject:
PLANT NAME
DOCKET NO. 50-
APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING REVISION OF
CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY, CLARIFICATION OF SRM
INSERT CONTROL ROD ACTION, AND A CLARIFICATION OF A FREQUENCY
EXAMPLE USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS.
Gentleman: In accordance with the provisions of 10 CFR 50.90
[LICENSEE] is submitting a request for an amendment to the technical
specifications (TS) for [PLANT NAME, UNIT NOS.].
The proposed amendment would: (1) Revise the TS surveillance
requirement (SR) frequency in TS 3.1.3, ``Control Rod OPERABILITY'',
(2) clarify the requirement to fully insert all insertable control
rods for the limiting condition for operation (LCO) in TS 3.3.1.2,
required Action E.2, ``Source Range Monitoring Instrumentation,''
and (3) revise Example 1.4-3 in Section 1.4 ``Frequency'' to clarify
the applicability of the 1.25 surveillance test interval extension.
Attachment 1 provides a description of the proposed change, the
requested confirmation of applicability, and plant-specific
verifications. Attachment 2 provides the existing TS pages marked up
to show the proposed change. Attachment 3 provides revised (clean)
TS pages. Attachment 4 provides a summary of the regulatory
commitments made in this submittal.
[LICENSEE] requests approval of the proposed License Amendment
by [DATE], with the amendment being implemented [BY DATE OR WITHIN X
DAYS].
In accordance with 10 CFR 50.91, a copy of this application,
with attachments, is being provided to the designated [STATE]
Official.
I declare under penalty of perjury under the laws of the United
Stats of America that I am authorized by [LICENSEE] to make this
request and that the foregoing is true and correct. (Note that
request may be notarized in lieu of using this oath or affirmation
statement). If you should have any questions regarding this
submittal, please contact [NAME, TELEPHONE NUMBER]
Sincerely,
[Name, Title]
Attachments:
1. Description and Assessment.
2. Proposed Technical Specification Changes.
3. Revised Technical Specification Pages.
4. Regulatory Commitments.
5. Proposed Technical Specification Bases Changes.
CC: NRC Project Manager.
NRC Regional Office.
NRC Resident Inspector.
State Contact.
Attachment 1--Description and Assessment
1.0 Description
The proposed amendment would: (1) Revise the TS surveillance
requirement (SR 3.1.3.2) frequency in TS 3.1.3, ``Control Rod
OPERABILITY'', (2) clarify the requirement to fully insert all
insertable control rods for the limiting condition for operation (LCO)
in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring
Instrumentation'', and (3) revise Example 1.4-3 in Section 1.4
``Frequency'' to clarify the applicability of the 1.25 surveillance
test interval extension.
The changes are consistent with Nuclear Regulatory Commission (NRC)
approved Industry/Technical Specification Task Force (TSTF) STS change
TSTF-475, Revision 1. The Federal Register notice published on [DATE]
announced the availability of this TS improvement through the
consolidated line item improvement process (CLIIP).
2.0 Assessment
2.1 Applicability of Published Safety Evaluation
[LICENSEE] has reviewed the safety evaluation dated [DATE] as part
of the CLIIP. This review included a review of the NRC staff's
evaluation, as well as the supporting information provided to support
TSTF-475, Revision 1. [LICENSEE] has concluded that the justifications
presented in the TSTF proposal and the safety evaluation prepared by
the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this
amendment for the incorporation of the changes to the [PLANT] TS.
2.2 Optional Changes and Variations
[LICENSEE] is not proposing any variations or deviations from the
TS changes described in the modified TSTF-475, Revision 1 and the NRC
staff's model safety evaluation dated [DATE].
3.0 Regulatory Analysis
3.1 No Significant Hazards Consideration Determination
[LICENSEE] has reviewed the proposed no significant hazards
consideration determination (NSHCD) published in the Federal Register
as part of the CLIIP. [LICENSEE] has concluded that the proposed NSHCD
presented in the Federal Register notice is applicable to [PLANT] and
is hereby incorporated by reference to satisfy the requirements of 10
CFR 50.91(a).
3.2 Verification and Commitments
As discussed in the notice of availability published in the Federal
Register on [DATE] for this TS improvement, the [LICENSEE] verifies the
applicability of TSTF-475 to [PLANT], and commits to establishing
Technical Specification Bases for TS as proposed in TSTF-475, Revision
1.
These changes are based on TSTF change traveler TSTF-475 (Revision
1) that proposes revisions to the BWR STS by: (1) Revising the
frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod,
from ``7 days after the control rod is withdrawn and THERMAL POWER is
greater than the LPSP of RWM'' to ``31 days after the control rod is
withdrawn and THERMAL POWER is greater than the LPSP of the RWM'', (2)
adding the word ``fully'' to LCO 3.3.1.2 Required Action E.2 (NUREG-
1434 only) to clarify the requirement to fully insert all insertable
control rods in core cells containing one or more fuel assemblies when
the associated SRM instrument is inoperable, and (3) revising Example
1.4-3 in Section 1.4 ``Frequency'' to clarify that the 1.25
surveillance test interval extension in SR 3.0.2 is applicable to time
periods discussed in NOTES in the ``SURVEILLANCE'' column in addition
to the time periods in the ``FREQUENCY'' column.
4.0 Environmental Evaluation
[LICENSEE] has reviewed the environmental evaluation included in
the model safety evaluation dated [DATE] as part of the CLIIP.
[LICENSEE]
[[Page 46108]]
has concluded that the staff's findings presented in that evaluation
are applicable to [PLANT] and the evaluation is hereby incorporated by
reference for this application.
Attachment 2--Proposed Technical Specification Changes (Mark-Up)
Attachment 3--Proposed Technical Specification Pages
Attachment 4--List of Regulatory Commitments
The following table identifies those actions committed to by
[LICENSEE] in this document. Any other statements in this submittal are
provided for information purposes and are not considered to be
regulatory commitments. Please direct questions regarding these
commitments to [CONTACT NAME].
------------------------------------------------------------------------
Regulatory commitments Due date/event
------------------------------------------------------------------------
[LICENSEE] will establish the Technical [Complete, implemented
Specification Bases for [TS B 3.1.3, TS B with amendment or within
3.1.4, and TS B 3.3.1.2] as adopted with the X days of implementation
applicable license amendment. of amendment].
[LICENSEE] will establish the water quality [Complete, implemented
controls as recommended by SIL No. 148, with amendment or within
Water Quality Control for the Control Rod X days of implementation
System,'' September 15, 1975. of amendment].
------------------------------------------------------------------------
Attachment 5--Proposed Changes to Technical Specification Bases Pages
Proposed No Significant Hazards Consideration Determination
Description of Amendment Request: [Plant Name] requests adoption of
an approved change to the Standard Technical Specifications (STS) for
General Electric (GE) Plants (NUREG-1433, BWR/4 and NUREG-1434, BWR/6)
and plant specific technical specifications (TS), that allows: (1)
Revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn
control rod, from ``7 days after the control rod is withdrawn and
THERMAL POWER is greater than the LPSP of RWM'' to ``31 days after the
control rod is withdrawn and THERMAL POWER is greater than the LPSP of
the RWM'', (2) adding the word ``fully'' to LCO 3.3.1.2 Required Action
E.2 (NUREG-1434 only) to clarify the requirement to fully insert all
insertable control rods in core cells containing one or more fuel
assemblies when the associated SRM instrument is inoperable, and (3)
revising Example 1.4-3 in Section 1.4 ``Frequency'' to clarify that the
1.25 surveillance test interval extension in SR 3.0.2 is applicable to
time periods discussed in NOTES in the ``SURVEILLANCE'' column in
addition to the time periods in the ``FREQUENCY'' column. The staff
finds that the proposed STS changes are acceptable because the number
of control rod manipulations is reduced thereby reducing the
opportunity for potential reactivity events while having a very minimal
impact on the extremely high reliability of the CRD system as discussed
in the technical evaluation section of this safety evaluation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM Insert Control Rod
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and NUREG-
1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency for
surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY'', (2) clarify the requirement to fully insert all
insertable control rods for the limiting condition for operation (LCO)
in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. The consequences of an accident
after adopting TSTF-475, Revision 1 are no different than the
consequences of an accident prior to adoption. Therefore, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. The proposed
change will not introduce new failure modes or effects and will not, in
the absence of other unrelated failures, lead to an accident whose
consequences exceed the consequences of accidents previously analyzed.
Thus, this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2 frequency
in TS 3.1.3, ``Control Rod OPERABILITY'', (2) clarify the requirement
to fully insert all insertable control rods for the limiting condition
for operation (LCO) in TS 3.3.1.2, ``Source Range Monitoring
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. The GE Nuclear Energy Report,
``CRD Notching Surveillance Testing for Limerick Generating Station,''
dated November 2006, concludes that extending the control rod notch
test interval from weekly to monthly is not expected to impact the
reliability of the scram system and that the analysis supports the
decision to change the surveillance frequency. Therefore, the proposed
changes in TSTF-475, Revision 1 are acceptable and do not involve a
significant reduction in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Dated at Rockville, Maryland, this 9th day of August, 2007.
[[Page 46109]]
For the Nuclear Regulatory Commission.
Carl Schulten,
Acting Chief, Technical Specifications Branch, Division of Inspection &
Regional Support, Office of Nuclear Reactor Regulation.
[FR Doc. E7-16138 Filed 8-15-07; 8:45 am]
BILLING CODE 7590-01-P
