[Federal Register Volume 72, Number 218 (Tuesday, November 13, 2007)]
[Notices]
[Pages 63935-63942]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-22159]
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NUCLEAR REGULATORY COMMISSION
Notice of Availability of Model Application Concerning 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: Notice of availability.
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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 revision of Standard Technical Specifications (STS),
NUREG-1430 (B&W), NUREG-1431 (Westinghouse), NUREG-1432 (CE), NUREG-
1433 (BWR/4) and NUREG-1434 (BWR/6). Specifically the SE addresses: (1)
The revision of the technical specification (TS) surveillance
requirement (SR) 3.1.3.2 frequency in STS 3.1.3, ``Control Rod
OPERABILITY,'' (NUREG-1433 and NUREG-1434), (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 (NUREG-1430 through NUREG-1434). 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 is 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 can request amendments, confirming the applicability of the SE
and NSHC determination to their plant licensing basis.
DATES: The NRC staff issued a Federal Register notice (72 FR 46103;
August 16, 2007) which provided a model SE, model application, and
model NSHC related to BWR plant control rod notch surveillance
frequency, BWR SRM control rod insertion action, and clarification of a
surveillance frequency example for all plant types. Similarly, the NRC
staff herein provides a revised model SE, model LAR, and model NSHC
incorporating changes based upon the public comments received. The NRC
staff can most efficiently consider applications based upon the model
LAR, which references the model SE, if the LAR is submitted within one
year of this Federal Register Notice.
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 standard technical specifications (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 following a preliminary assessment by the NRC staff
and finding that the change will likely be offered for adoption by
licensees. 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 to proceed with announcing the availability of the change for
proposed
[[Page 63936]]
adoption by licensees. Those licensees opting to apply for the subject
change to technical specifications 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 BWR TS control rod SR
testing frequency, clarification of BWR TS control insertion
requirements, and clarification of SR frequency discussions for all
pant types. 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 http://www.nrc.gov/reactors/operating/licensing/techspecs.html.
*** Reviewer's Note ***
TSTF-475 involves three changes to the Standard Technical
Specifications NUREGs that, depending upon the adopting plant, may
or may not be adopted by a plant. The first changes the surveillance
frequency for control rod notch testing from 7 to 31 days, and
applies to BWR/4 and BWR/6 plants (NUREG-1433 & NUREG-1434). The
second adds the word ``fully'' to a Required Action statement to
clarify that control rods should be fully inserted, and applies to
only the BWR/6 plants (NUREG-1434). The third change clarifies the
usage of the 1.25 surveillance frequency interval extension, and
applies to all plants (NUREG-1430 through NUREG-1434). The model
application and model safety evaluation will need to be tailored
(where brackets indicate) for plant specific applications.
Applicability
This proposed TS change modifies TS control rod SR testing
frequency and clarifies TS control insertion requirements for BWR
plants, and clarifies SR frequency discussions for all NSSS plant
types. 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 the review.
To efficiently process the incoming license amendment applications,
the staff requests that each licensee applying for the changes proposed
in TSTF-475, Revision 1, include TS Bases for the proposed TS
consistent with the TS Bases proposed in TSTF-475, Revision 1 (note:
the change to STS Section 1.4 does not entail a Bases change). The
staff is requesting that the TS Bases be included with the proposed
license amendments in this case because the changes to the TS and the
changes to the associated TS Bases form an integral change to a plant's
licensing basis. To ensure that the overall change, including the TS
Bases, includes appropriate regulatory controls, the staff plans to
condition the issuance of each license amendment on the licensee's
incorporation of the changes into the TS Bases document and that the
licensee control changes to the TS Bases in accordance with the
licensees TS Bases Control Program. The CLIIP does not prevent
licensees from requesting an alternative approach or proposing the
changes without the requested TS Bases. However, deviations from the
approach recommended in this notice may require additional review by
the NRC staff and may increase the time and resources needed for the
review. Significant variations from the approach, or inclusion of
additional changes to the license, will result in staff rejection of
the submittal. Instead, licensees desiring significant variations and/
or additional changes should submit a LAR that does not request to
adopt TSTF-475, Revision 1, under CLIIP.
Public Notices
The staff issued a Federal Register Notice (72 FR 46103, August 16,
2007) that requested public comment on the NRC's pending action to
approve the modification of BWR TS control rod SR testing frequency,
clarification of BWR TS control insertion requirements, and
clarification of SR frequency discussions for all pant types, as
proposed in TSTF-475, Revision 1. The TSTF-475, Revision 1, can be
viewed on the NRC's web page at http://www.nrc.gov/reactors/operating/licensing/techspecs.html. TSTF-475, Revision 1, may be examined, and/or
copied for a fee, at the NRC's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records are accessible electronically from
the ADAMS Public Library component on the NRC Web site, (the Electronic
Reading Room) at http://www.nrc.gov/reading-rm/adams.html.
In response to the notice soliciting comments from interested
members of the public about the modification of BWR TS control rod SR
testing frequency, clarification of BWR TS control insertion
requirements, and clarification of SR frequency discussions for all
pant types, the staff received one set of comments (from the TSTF
Owners Groups, representing licensees). The specific comments are
provided and discussed below:
1. Comment: TSTF-475 contains three changes: The revision to SR
3.1.3.2 which is applicable to NUREG-1433 and NUREG-1434 (the Improved
Standard Technical Specifications, or ISTS, for BWR/4 and BWR/6
plants), the change to Specification 3.3.1.2, Required Action E.2 which
is applicable to NUREG-1434 (the ISTS for BWR/6 plants), and the change
to Example 1.4-3 which is applicable to NUREG-1430 through -1434 (the
ISTS for all plant types). The applicability of the third change to all
plant types is clearly indicated on the Traveler cover page and in the
justification (last paragraph of Section 2.0, ``Proposed Change.'')
However, the Notice for Comment, model Safety Evaluation, model
application, and No Significant Hazards Considerations Determination
(NSHC) incorrectly state that TSTF-475 is only applicable to BWR
plants.
The Notice, the model application, model Safety Evaluation, and
NSHC should be revised to state that the change to Example 1.4-3 is
applicable to all plant types. The model Safety Evaluation, model
application, and NSHC should be revised to bracket (e.g., indicate as
optional) the BWR/4 and BWR/6 specific changes so that the documents
are applicable to a BWR/6 plant adopting all three changes, a BWR/4
plant adopting the SR 3.1.3.2 and Example 1.4-3 changes, or a
pressurized water reactor (PWR) plant adopting only the Example 1.4-3
change.
Response: The staff agrees with the comment and the model
application, model Safety Evaluation, and NSHC have been revised
accordingly.
2. Comment: In Section 3.0, ``Technical Evaluation,'' of the
Notice, reference is made three times to the ``BWROG TSTF'' or ``BWROG
TSTF-475.'' The Technical Specifications Task Force (TSTF) is sponsored
by the Boiling Water Reactor Owners Group and the Pressurized Water
Reactor Owners Group. The proper designation is either ``TSTF'' or
``Owners Group TSTF.''
Response: The staff agrees with the comment and Section 3.0 of the
model Safety Evaluation has been revised by removing explicit reference
to the BWROG in referring to TSTF-475.
3. Comment: In Section 3.0, ``Technical Evaluation,'' the model
Safety Evaluation states, ``Therefore, the NRC staff finds the change
acceptable
[[Page 63937]]
with the commitment to implement GE water quality for the CRD system
recommendations.'' In the model application, a regulatory commitment is
included which states, ``[LICENSEE] will establish the water quality
controls as recommended by SIL No. 148, Water Quality Control for the
Control Rod System,'' September 15, 1975.'' This commitment should be
removed.
The TSTF's justification for TSTF-475 made no mention of and did
not rely on water quality controls. The TSTF's July 3, 2006 response to
the NRC's March 21, 2003 Request for Additional Information (RAI) did
not credit water chemistry controls. As stated in the justification and
the Staff's model Safety Evaluation, 30 years of operating experience
at BWRs without a control rod drive failure detected by the weekly
notch testing is sufficient to demonstrate the acceptability of the
change.
The reference is technically incorrect. Supplement 1 to SIL No. 148
was issued in June 2004 and updates the SIL to bring it into alignment
with current Electric Power Research Institute (EPRI) BWR water
chemistry requirements, which were in conflict with the 1975 version of
SIL.
The NRC's Technical Evaluation in the draft Safety Evaluation did
not reference SIL No. 148 (either the 1975 version or the current
version).
It is not appropriate for the NRC to require commitments to
documents that were not relied on in the licensee's application, were
not reviewed by the NRC, and were not discussed in the NRC's technical
evaluation. Therefore, the reference to water chemistry controls in the
model Safety Evaluation and the commitment in the model application
should be removed.
Response: The staff agrees with the comment and the requirements
for a commitment to establish water quality controls as recommended by
SIL No. 148, Water Quality Control for the Control Rod System, in the
model Safety Evaluation and in the model application have been removed.
4. Comment: Model Application: Attachment 5, ``Proposed Technical
Specification Bases,'' should be marked as optional. There are no Bases
changes associated with the PWR-applicable changes to Section 1.4.
Furthermore, the Bases changes associated with TSTF-475 simply reflect
the changes made to the specifications. It should be left to the
licensee whether to submit Bases changes with the amendment request.
The third paragraph omits Attachment 5, which is shown in the list of
attachments below the signature. Attachment 3, ``Proposed Technical
Specification Pages,'' should also be marked as optional as not all
licensee's submit retyped Technical Specification pages as attachments
to their amendment requests.
Response: The staff does not agree with the comment. For those
sections of the technical specifications that are changed in accordance
with TSTF-475 and that have Bases, the Bases must be changed to reflect
the change in accordance with TSTF-475. TS Section 1.4, that does not
have Bases, does not need to have Bases changes submitted, and for
those plants that are only adopting the TS Section 1.4 change, the
Model Application Attachment 5, ``Proposed Technical Specification
Bases,'' will be revised to indicate that the submittal of revised
Bases pages is optional in that case. The staff does not see a need to
revise Model Application Attachment 3. The staff expects to see the
licensee's Bases changes associated with the adoption of TSTF-475.
5. Comment: Model Application: The Model Application states, ``I
declare under penalty of perjury under the laws of the United States of
America that I am authorized by [LICENSEE] to make this request and
that the foregoing is true and correct.'' This statement is not
consistent with the recommended statement given in RIS 2001-18,
``Requirements for Oath or Affirmation.'' RIS 2001-18 recommends the
statement, ``I declare [or certify, verify, state] under penalty of
perjury that the foregoing is true and correct.'' Note that RIS 2001-18
states that this statement must be used verbatim. We recommend that the
Model Application be revised to be consistent with RIS 2001-18.
Response: The staff agrees with the comment and the requirement in
the model application for oath or affirmation has been reworded to be
consistent with RIS 2001-18.
6. Comment: Attachment 4: The regulatory commitment states
``[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.'' This statement is incorrect as the Bases changes
included for information with the license amendment request are not
``adopted'' with the license amendment. Bases changes are made under
licensee control under the Technical Specification Bases Control
Program. We recommend revising the commitment to state ``[LICENSEE]
will implement Technical Specification Bases for TS [3.1.3, 3.1.4, and
3.3.1.2] consistent with those shown in TSTF-475, Revision 1, ``Control
Rod Notch Testing Frequency and SRM Insert Control Rod Action.'' The
commitment should also be marked as optional consistent with Comments 1
and 4, as the PWR-applicable change to Section 1.4 has no associated
Bases changes.
Response: The staff agrees with the comment in the sense that the
Bases are not adopted as a license amendment is adopted, and therefore
the wording of the commitment will be revised to state, ``[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] consistent with those shown in TSTF-475,
Revision 1, ``Control Rod Notch Testing Frequency and SRM Insert
Control Rod Action.'' The staff does not agree with the comment with
respect to the Bases being provided purely for information and that the
commitment is optional. The staff will review the Bases changes to
ensure they are acceptable. If a licensee is only adopting the TS
Section 1.4 portion of the TSTF-475 change, then the commitment would
not apply, otherwise it would apply.
7. Comment: Model NSHC: To be consistent with 10 CFR 50.91(a), the
title of Criterion 2 should be revised to add the word ``Accident''
before ``Previously Evaluated.'' Specifically, it should state, ``The
Proposed Change Does Not Create the Possibility of a New or Different
Kind of Accident from any Accident Previously Evaluated.''
Response: The staff agrees with the comment and the model NSHC
Criterion 2 statement has been reworded accordingly.
For the Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 5th day of November, 2007.
Timothy J. Kobetz,
Chief, Technical Specifications Branch, Division of Inspection and
Regional Support, Office of Nuclear Reactor Regulation.
Model Safety Evaluation, U.S. Nuclear Regulatory Commission, Office of
Nuclear Reactor Regulation, Consolidated Line Item Improvement,
Technical Specification Task Force (TSTF) Change TSTF-475, Revision 1,
Control Rod Notch Testing Frequency, Source Range Monitor Technical
Specification Action to Insert Control Rods, and Surveillance Frequency
Discussions
1.0 Introduction
By letter dated August 30, 2004, the TSTF submitted a request
(Reference 1) for changes to the Standard Technical Specifications
(STS): NUREG-1430 Standard Technical Specifications B&W Plants
(Reference 2); NUREG-1431 Standard Technical Specifications
Westinghouse Plants (Reference 3); NUREG-1432 Standard Technical
[[Page 63938]]
Specifications Combustion Engineering Plants (Reference 4); NUREG-1433,
Standard Technical Specifications General Electric Plants, BWR/4
(Reference 5); and NUREG-1434, Standard Technical Specifications
General Electric Plants, BWR/6 (Reference 6). The proposed changes
would: (1) Revise the TS control rod notch surveillance frequency in TS
3.1.3, ``Control Rod OPERABILITY,'' (NUREG-1433 and NUREG-1434), (2)
clarify the TS requirement for inserting control rods for one or more
inoperable SRMs in MODE 5 (NUREG-1434 only), and (3) revise one Example
in Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension (NUREG-1430 through NUREG-1434).
These changes are based on Technical Specifications Task Force
(TSTF) change traveler TSTF-475, Revision 1, that proposes revisions to
the reference 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'' (NUREG-1433 and NUREG-1434), (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 (NUREG-1430 through
NUREG-1434).
[The purpose of the 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.]
The purpose of the change to Example 1.4-3 in Section 1.4
``Frequency'' is to clarify the applicability of the 25% allowance of
SR 3.0.2 to time periods discussed in NOTES in the ``SURVEILLANCE''
column as well as to time periods in the ``FREQUENCY'' column.
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 TSTF to justify the submitted license
amendment request to [revise the weekly control rod notch frequency to
monthly (STS NUREG-1433 and NUREG-1434)], [clarify the SRM TS action
for inserting control rods (NUREG-1434 only), and] revise the
discussion of 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 (Reference 1)--Provided a
description of the proposed changes in TSTF-475 that changes 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.
[TSTF letter TSTF-06-13 (Reference 8)--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.
BWROG letter BWROG-06036 (Reference 9)--Provided the GE
Nuclear Energy Report, ``CRD Notching Surveillance Testing for Limerick
Generating Station,'' in which CRD notching frequency and CRD
performance were evaluated.
TSTF letter TSTF-07-19 (Reference 10)--Provided response
to NRC staff RAI on CRD performance in Control Cell Core (CCC) designed
plants, including TSTF-475, Revision 1.
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
[[Page 63939]]
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 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 TSTF-475 proposal to amend the
(NUREG-1433 and NUREG-1434) TS SR 3.1.3.2, ``Control Rod OPERABILTY''
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. The utilities should
consider the replacement of the CRT, when possible, with the GE CRT
improved design.
The NRC staff has reviewed the 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
[[Page 63940]]
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 TSTF-475 proposal to amend (NUREG-
1430 through NUREG-1434) 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 [(NUREG-1433 and NUREG-1434) TS sections SR 3.1.3.2, ``Control
Rod OPERABILTY,'' (NUREG-1434) LCO 3.3.1.2 Required Action E.2,
``Source Range Monitor (SRM) Instrumentation,'' and] (NUREG-1430
through NUREG-1434) 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, and] will clarify the
1.25 provision in SR 3.0.2. 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. 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 Action,'' August 30, 2004, ADAMS accession
number ML042520035.
2. NUREG-1430, ``Standard Technical Specifications Babcock and
Wilcox Plants, Revision 3,'' August 31, 2003.
3. NUREG-1431, ``Standard Technical Specifications Westinghouse
Plants, Revision 3,'' August 31, 2003.
4. NUREG-1432, ``Standard Technical Specifications Combustion
Engineering Plants, Revision 3,'' August 31, 2003.
5. NUREG-1433, ``Standard Technical Specifications General Electric
Plants, BWR/4, Revision 3,'' August 31, 2003.
6. NUREG-1434, ``Standard Technical Specifications General Electric
Plants, BWR/6, Revision 3,'' August 31, 2003.
7. Letter TSTF-07-19, Response from the Technical Specifications
Task Force to the NRC, ``Request for Additional Information (RAI)
Regarding TSTF-475 Revision 0,'' Control Rod Notch Testing Frequency
and SRM Insert Control Rod Action,'' dated February 28, 2007, (TSTF-475
Revision 1 is an enclosure), ADAMS accession number ML071420428.
8. 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.
9. 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, with Enclosure of the GE
Nuclear Energy Report, ``CRD Notching Surveillance Testing for Limerick
Generating Station,'' dated November 2006, ADAMS accession number
ML063250258.
10. 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 ML071420428].
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 Regular 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
[[Page 63941]]
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 [or certify, verify, state] under penalty of perjury
that the foregoing is true and correct.
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 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 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] 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 with
Specification Bases for [TS B 3.1.3, TS B amendment OR within X days
3.1.4, and TS B 3.3.1.2] consistent with of implementation of
those shown in TSTF-475, Revision 1, amendment].
``Control Rod Notch Testing Frequency and
SRM Insert Control Rod Action.''].
------------------------------------------------------------------------
[[Page 63942]]
ATTACHMENT 5--PROPOSED CHANGES TO TECHNICAL SPECIFICATION BASES PAGES
[Not required for plants only adopting portion of TSTF-475 change
pertaining to TS Section 1.4 that provides example to SR Frequency]
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 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 and] the discussion of the
SR Frequency example provides clarification.
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 Accident 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,'' 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 5th day of November, 2007.
For the Nuclear Regulatory Commission.
Timothy J. Kobetz,
Section Chief, Technical Specifications Branch, Division of Inspection
& Regional Support, Office of Nuclear Reactor Regulation.
[FR Doc. E7-22159 Filed 11-9-07; 8:45 am]
BILLING CODE 7590-01-P