[Federal Register Volume 69, Number 226 (Wednesday, November 24, 2004)]
[Notices]
[Pages 68412-68420]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-26008]
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NUCLEAR REGULATORY COMMISSION
Notice of Opportunity To Comment on Model Safety Evaluation on
Technical Specification Improvement To Modify Requirements Regarding
the Addition of LCO 3.0.8 on the Inoperability of Snubbers Using the
Consolidated Line Item Improvement Process
AGENCY: Nuclear Regulatory Commission.
ACTION: Request for comment.
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SUMMARY: Notice is hereby given that the staff of the Nuclear
Regulatory Commission (NRC) has prepared a model safety evaluation (SE)
relating to the impact of inoperable non-technical specification
snubbers on supported systems in technical specifications (TS). The NRC
staff has also prepared a model no-significant-hazards-consideration
(NSHC) determination relating to this matter. The purpose of these
models is to permit the NRC to efficiently process amendments that
propose to add an LCO 3.0.8 that provides a delay time for entering a
supported system TS when the inoperability is due solely to an
inoperable snubber, if risk is assessed and managed. Licensees of
nuclear power reactors to which the models apply could then request
amendments, confirming the applicability of the SE and NSHC
determination to their reactors. The NRC staff is requesting comment on
the model SE and model NSHC determination prior to announcing their
availability for referencing in license amendment applications.
DATES: The comment period expires December 27, 2004. Comments received
after this date will be considered if it is practical to do so, but the
Commission is able to ensure consideration only for comments received
on or before this date.
ADDRESSES: Comments may be submitted either electronically or via U.S.
mail. Submit written comments to Chief, Rules and Directives Branch,
Division of Administrative Services, Office of Administration, Mail
Stop: T-6 D59, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001. Hand deliver comments to: 11545 Rockville Pike, Rockville,
Maryland, between 7:45 a.m. and 4:15 p.m. on Federal workdays. Copies
of comments received may be examined at the NRC's Public Document Room,
11555 Rockville Pike (Room O-1F21), Rockville, Maryland. Comments may
be submitted by electronic mail to [email protected].
FOR FURTHER INFORMATION CONTACT: Tom Boyce, Mail Stop: O-12H4, Division
of Inspection Program Management, Office of Nuclear Reactor Regulation,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
telephone 301-415-0184.
SUPPLEMENTARY INFORMATION:
Background
Regulatory Issue Summary 2000-06, ``Consolidated Line Item
Improvement Process for Adopting Standard Technical Specification
Changes for Power Reactors,'' was issued on March 20, 2000. The
consolidated line item improvement process (CLIIP) is intended to
improve the efficiency of NRC licensing processes by processing
proposed changes to the standard technical specifications (STS) in a
manner that supports subsequent license amendment applications. The
CLIIP includes an opportunity for the public to comment on a proposed
change to the STS after a preliminary assessment by the NRC staff and a
finding that the change will likely be offered for adoption by
licensees. This notice solicits comment on a proposed change that
allows a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed. The CLIIP directs the NRC staff to evaluate any
comments received for a proposed change to the STS and to either
reconsider the change or announce the availability of the change for
adoption by licensees. 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 addition of LCO 3.0.8 to the TS which
provides a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed. 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-372. TSTF-372 can be viewed on the NRC's Web page at http://www.nrc.gov/reactors/operating/licensing/techspecs.html.
Applicability
This proposal to modify technical specification requirements by the
[[Page 68413]]
addition of LCO 3.0.8, as proposed in TSTF-372, is applicable to all
licensees who have adopted or will adopt, in conjunction with the
proposed change, technical specification requirements for a Bases
control program consistent with the TS Bases Control Program described
in Section 5.5 of the applicable vendor's STS.
To efficiently process the incoming license amendment applications,
the staff requests that each licensee applying for the changes proposed
in TSTF-372 include Bases for the proposed TS consistent with the Bases
proposed in TSTF-372. In addition, licensees that have not adopted
requirements for a Bases control program by converting to the improved
STS or by other means are requested to include the requirements for a
Bases control program consistent with the STS in their application for
the proposed change. The need for a Bases control program stems from
the need for adequate regulatory control of some key elements of the
proposal that are contained in the proposed Bases for LCO 3.0.8. The
staff is requesting that the Bases be included with the proposed
license amendments in this case because the changes to the TS and the
changes to the associated Bases form an integral change to a plant's
licensing basis. To ensure that the overall change, including the
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 Bases document and on requiring
the licensee to control the changes in accordance with the Bases
Control Program. The CLIIP does not prevent licensees from requesting
an alternative approach or proposing the changes without the requested
Bases and Bases control program. 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.
Public Notices
This notice requests comments from interested members of the public
within 30 days of the date of publication in the Federal Register.
After evaluating the comments received as a result of this notice, the
staff will either reconsider the proposed change or announce the
availability of the change in a subsequent notice (perhaps with some
changes to the safety evaluation or the proposed no significant hazards
consideration determination as a result of public comments). If the
staff announces the availability of the change, licensees wishing to
adopt the change must submit an application in accordance with
applicable rules and other regulatory requirements. For each
application the staff will publish a notice of consideration of
issuance of amendment to facility operating licenses, a proposed no
significant hazards consideration determination, and a notice of
opportunity for a hearing. The staff will also publish a notice of
issuance of an amendment to an operating license to announce the
modification of requirements for mode change limitations for each plant
that receives the requested change.
Proposed Safety Evaluation
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor
Regulation, Consolidated Line Item Improvement, Technical Specification
Task Force (TSTF) Change TSTF-372; The Addition of Limiting Condition
for Operation (LCO) 3.0.8 on the Inoperability of Snubbers
1.0 Introduction
On April 23, 2004, the Nuclear Energy Institute (NEI) Risk Informed
Technical Specifications Task Force (RITSTF) submitted a proposed
change, TSTF-372, Revision 4, to the standard technical specifications
(STS) (NUREGs 1430-1434) on behalf of the industry (TSTF-372, Revisions
1 through 3 were prior draft iterations). TSTF-372, Revision 4, is a
proposal to add an STS Limiting Condition for Operation (LCO) 3.0.8,
allowing a delay time for entering a supported system technical
specification (TS), when the inoperability is due solely to an
inoperable snubber, if risk is assessed and managed. The postulated
seismic event requiring snubbers is a low-probability occurrence and
the overall TS system safety function would still be available for the
vast majority of anticipated challenges.
This proposal is one of the industry's initiatives being developed
under the risk-informed technical specifications program. These
initiatives are intended to maintain or improve safety through the
incorporation of risk assessment and management techniques in TS, while
reducing unnecessary burden and making technical specification
requirements consistent with the Commission's other risk-informed
regulatory requirements, in particular the Maintenance Rule.
The proposed change adds a new limiting condition of operation, LCO
3.0.8, to the TS. LCO 3.0.8 allows licensees to delay declaring an LCO
not met for equipment, supported by snubbers unable to perform their
associated support functions, when risk is assessed and managed. This
new LCO 3.0.8 states:
When one or more required snubbers are unable to perform their
associated support function(s), any affected supported LCO(s) are
not required to be declared not met solely for this reason if risk
is assessed and managed, and:
a. The snubbers not able to perform their associated support
function(s) are associated with only one train or subsystem of a
multiple train or subsystem supported system or are associated with
a single train or subsystem supported system and are able to perform
their associated support function within 72 hours; or
b. The snubbers not able to perform their associated support
function(s) are associated with more than one train or subsystem of
a multiple train or subsystem supported system and are able to
perform their associated support function within 12 hours.
At the end of the specified period the required snubbers must be
able to perform their associated support function(s), or the
affected supported system LCO(s) shall be declared not met.
2.0 Regulatory Evaluation
In 10 CFR 50.36, the Commission established its regulatory
requirements related to the content of TS. Pursuant to 10 CFR 50.36, TS
are required to include items in the following five specific categories
related to station operation: (1) Safety limits, limiting safety system
settings, and limiting control settings; (2) limiting conditions for
operation (LCOs); (3) surveillance requirements (SRs); (4) design
features; and (5) administrative controls. The rule does not specify
the particular requirements to be included in a plant's TS. As stated
in 10 CFR 50.36(c)(2)(i), the ``Limiting conditions for operation are
the lowest functional capability or performance levels of equipment
required for safe operation of the facility. When a limiting condition
for operation of a nuclear reactor is not met, the licensee shall shut
down the reactor or follow any remedial action permitted by the
technical specification * * *.'' TS Section 3.0, on ``LCO and SR
Applicability,'' provides details or ground rules for complying with
the LCOs. Snubbers are chosen in lieu of rigid supports in areas where
restricting thermal growth during normal operation would induce
excessive stresses in the piping nozzles or other equipment. Although
they are classified as component standard supports, they are not
designed to provide any transmission of force during normal plant
operations. However, in the presence of dynamic transient loadings,
which are induced by seismic events as well as by plant accidents and
transients, a snubber functions as a rigid support. The location and
size of the
[[Page 68414]]
snubbers are determined by stress analysis based on different
combinations of load conditions, depending on the design classification
of the particular piping.
Prior to the conversion to the improved STS, TS requirements
applied directly to snubbers. These requirements included:
A requirement that snubbers be functional and in service
when the supported equipment is required to be operable,
A requirement that snubber removal for testing be done
only during plant shutdown,
A requirement that snubber removal for testing be done on
a one-at-a-time basis when supported equipment is required to be
operable during shutdown,
A requirement to repair or replace within 72 hours any
snubbers, found to be inoperable during operation in Modes 1 through 4,
to avoid declaring any supported equipment inoperable,
A requirement that each snubber be demonstrated operable
by periodic visual inspections, and
A requirement to perform functional tests on a
representative sample of at least 10% of plant snubbers, at least once
every 18 months during shutdown.
In the late 1980s, a joint initiative of the NRC and industry was
undertaken to improve the STS. This effort identified the snubbers as
candidates for relocation to a licensee-controlled document based on
the fact that the TS requirements for snubbers did not meet any of the
four criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the improved
STS. The NRC approved the relocation without placing any restriction on
the use of the relocated requirements. However, this relocation
resulted in different interpretations between the NRC and the industry
regarding its implementation. The NRC has stated, that since snubbers
are supporting safety equipment that is in the TS, the definition of
Operability must be used to immediately evaluate equipment supported by
a removed snubber and, if found inoperable, the appropriate TS required
actions must be entered. This interpretation has in practice eliminated
the 72-hour delay to enter the actions for the supported equipment that
existed prior to the conversion to the improved STS (the only exception
is if the supported system has been analyzed and determined to be
Operable without the snubber). The industry has argued that since the
NRC approved the relocation without placing any restriction on the use
of the relocated requirements, the licensee controlled document
requirements for snubbers should be invoked before the supported
system's TS requirements become applicable. The industry's
interpretation would, in effect, restore the 72-hour delay to enter the
actions for the supported equipment that existed prior to the
conversion to the improved STS. However, prior to the conversion to the
improved STS, the delay was applicable only to snubbers found to be
inoperable (i.e., to emergent conditions only). The industry's
interpretation would allow a time delay for all conditions, including
snubber removal for testing at power, that was not allowed prior to the
conversion to the improved STS. The option to relocate the snubbers to
a licensee controlled document, as part of the conversion to improved
STS, has resulted in non-uniform and inconsistent treatment of
snubbers. On the one hand, plants that have relocated snubbers from
their TS are allowed to change the TS requirements for snubbers under
the auspices of 10 CFR 50.59, but they are not allowed a 72-hour delay
before they enter the actions for the supported equipment. On the other
hand, plants that have not converted to improved STS have retained the
72-hour delay if snubbers are found to be inoperable, but they are not
allowed to use 10 CFR 50.59 to change TS requirements for snubbers. It
should also be noted that a few plants that converted to the improved
STS chose not to relocate the snubbers to a licensee-controlled
document and, thus, retained the 72-hour delay. In addition, it is
important to note that unlike plants that have not relocated, plants
that have relocated can perform functional tests on the snubbers at
power (as long as they enter the actions for the supported equipment)
and at the same time can reduce the testing frequency (as compared to
plants that have not relocated) if it is justified by 10 CFR 50.59
assessments. Some potential undesirable consequences of this
inconsistent treatment of snubbers are:
Performance of testing during crowded time period windows
when the supported system is inoperable with the potential to reduce
the snubber testing to a minimum since the relocated snubber
requirements are controlled by the licensee,
Performance of testing during crowded windows when the
supported system is inoperable with the potential to increase the
unavailability of safety systems, and
Performance of testing and maintenance on snubbers
affecting multiple trains of the same supported system during the 7
hours allotted before entering MODE 3 under LCO 3.0.3.
To remove the inconsistency in the treatment of snubbers among
plants, the TSTF proposed a risk-informed TS change that introduces a
delay time before entering the actions for the supported equipment,
when one or more snubbers are found inoperable or removed for testing,
if risk is assessed and managed. Such a delay time will provide needed
flexibility in the performance of maintenance and testing during power
operation and at the same time will enhance overall plant safety by:
Avoiding unnecessary unscheduled plant shutdowns and,
thus, minimizing plant transition and realignment risks,
Avoiding reduced snubber testing, and thus increasing the
availability of snubbers to perform their supporting function,
Performing most of the required testing and maintenance
during the delay time when the supported system is available to
mitigate most challenges and, thus, avoiding increases in safety system
unavailability, and
Providing explicit risk-informed guidance in areas in
which that guidance currently does not exist, such as the treatment of
snubbers impacting more than one redundant train of a supported system.
The proposed TS change is described in Sections 1.0 and 2.0. The
technical evaluation and approach used to assess its risk impact is
discussed in Section 3.0. The results and insights of the risk
assessment are presented and discussed in Section 3.1. Section 3.2
summarizes the staff's conclusions from the review of the proposed TS
change.
3.0 Technical Evaluation
The industry submitted TSTF-372, Revision 4, ``Addition of LCO
3.0.8, Inoperability of Snubbers'' in support of the proposed TS
change. This submittal (Ref. 1) documents a risk-informed analysis of
the proposed TS change. Probabilistic risk assessment (PRA) results and
insights are used, in combination with deterministic and defense-in-
depth arguments, to identify and justify delay times for entering the
actions for the supported equipment associated with inoperable snubbers
at nuclear power plants. This is in accordance with guidance provided
in Regulatory Guides (RGs) 1.174 and 1.177 (Refs. 2 and 3,
respectively).
The risk impact associated with the proposed delay times for
entering the TS actions for the supported equipment can be assessed
using the same approach as for allowed completion time (CT) extensions.
Therefore, the risk
[[Page 68415]]
assessment was performed following the three-tiered approach
recommended in RG 1.177 for evaluating proposed extensions in currently
allowed CTs:
The first tier involves the assessment of the change in
plant risk due to the proposed TS change. Such risk change is expressed
(1) by the change in the average yearly core damage frequency
([Delta]CDF) and the average yearly large early release frequency
([Delta]LERF) and (2) by the incremental conditional core damage
probability (ICCDP) and the incremental conditional large early release
probability (ICLERP). The assessed [Delta]CDF and [Delta]LERF values
are compared to acceptance guidelines, consistent with the Commission's
Safety Goal Policy Statement as documented in RG 1.174, so that the
plant's average baseline risk is maintained within a minimal range. The
assessed ICCDP and ICLERP values are compared to acceptance guidelines
provided in RG 1.177, which aim at ensuring that the plant risk does
not increase unacceptably during the period the equipment is taken out
of service.
The second tier involves the identification of potentially
high-risk configurations that could exist if equipment in addition to
that associated with the change were to be taken out of service
simultaneously, or other risk-significant operational factors such as
concurrent equipment testing were also involved. The objective is to
ensure that appropriate restrictions are in place to avoid any
potential high-risk configurations.
The third tier involves the establishment of an overall
configuration risk management program (CRMP) to ensure that potentially
risk-significant configurations resulting from maintenance and other
operational activities are identified. The objective of the CRMP is to
manage configuration-specific risk by appropriate scheduling of plant
activities and/or appropriate compensatory measures.
A simplified bounding risk assessment was performed to justify the
proposed addition of LCO 3.0.8 to the TS. This approach was
necessitated by (1) the general nature of the proposed TS changes
(i.e., they apply to all plants and are associated with an undetermined
number of snubbers that are not able to perform their function), (2)
the lack of detailed engineering analyses that establish the
relationship between earthquake level and supported system pipe failure
probability when one or more snubbers are inoperable, and (3) the lack
of seismic risk assessment models for most plants. The simplified risk
assessment is based on the following major assumptions, which the staff
finds acceptable, as discussed below:
The accident sequences contributing to the risk increase
associated with the proposed TS changes are assumed to be initiated by
a seismically-induced loss-of-offsite-power (LOOP) event with
concurrent loss of all safety system trains supported by the out-of-
service snubbers. In the case of snubbers associated with more than one
train (or subsystem) of the same system, it is assumed that all
affected trains (or subsystems) of the supported system are failed.
This assumption was introduced to allow the performance of a simple
bounding risk assessment approach with application to all plants. This
approach was selected due to the lack of detailed plant-specific
seismic risk assessments for most plants and the lack of fragility data
for piping when one or more supporting snubbers are inoperable.
The LOOP event is assumed to occur due to the seismically-
induced failure of the ceramic insulators used in the power
distribution systems. These ceramic insulators have a high confidence
(95%) of low probability (5%) of failure (HCLPF) of about 0.1g,
expressed in terms of peak ground acceleration. Thus, a magnitude 0.1g
earthquake is conservatively assumed to have 5% probability of causing
a LOOP initiating event. The fact that no LOOP events caused by higher
magnitude earthquakes were considered is justified because (1) the
frequency of earthquakes decreases with increasing magnitude and (2)
historical data (References 4 and 5) indicate that the mean seismic
capacity of ceramic insulators (used in seismic PRAs), in terms of peak
ground acceleration, is about 0.3g, which is significantly higher than
the 0.1g HCLPF value. Therefore, the simplified analysis, even though
it does not consider LOOP events caused by earthquakes of magnitude
higher than 0.1g, bounds a detailed analysis which would use mean
seismic failure probabilities (fragilities) for the ceramic insulators.
Analytical and experimental results obtained in the mid-
eighties as part of the industry's ``Snubber Reduction Program''
(References 4 and 6) indicated that piping systems have large margins
against seismic stress. The assumption that a magnitude 0.1g earthquake
would cause the failure of all safety system trains supported by the
out-of-service snubbers is very conservative because safety piping
systems could withstand much higher seismic stresses even when one or
more supporting snubbers are out of service. The actual piping failure
probability is a function of the stress allowable and the number of
snubbers removed for maintenance or testing. Since the licensee
controlled testing is done on only a small (about 10%) representative
sample of the total snubber population, it is not expected to have more
than a few snubbers supporting a given safety system out for testing at
a time. Furthermore, since the testing of snubbers is a planned
activity, licensees have flexibility in selecting a sample set of
snubbers for testing from a much larger population by conducting
configuration-specific engineering and/or risk assessments. Such a
selection of snubbers for testing provides confidence that the
supported systems would perform their functions in the presence of a
design-basis earthquake and other dynamic loads and, in any case, the
risk impact of the activity will remain within the limits of
acceptability defined in risk-informed RGs 1.174 and 1.177.
The analysis assumes that one train (or subsystem) of all
safety systems is unavailable during snubber testing or maintenance (an
entire system is assumed unavailable if a removed snubber is associated
with both trains of a two-train system). This is a very conservative
assumption for the case of corrective maintenance since it is unlikely
that a visual inspection will reveal that one or more snubbers across
all supported systems are inoperable. This assumption is also
conservative for the case of the licensee-controlled testing of
snubbers since such testing is performed only on a small representative
sample.
In general, no credit is taken for recovery actions and
alternative means of performing a function, such as the function
performed by a system assumed failed (e.g., when LCO 3.0.8b applies).
However, most plants have reliable alternative means of performing
certain critical functions. For example, feed and bleed (F&B) can be
used to remove heat in most pressurized water reactors (PWRs) when
auxiliary feedwater (AFW), the most important system in mitigating LOOP
accidents, is unavailable. Similarly, if high pressure makeup (e.g.,
reactor core isolation cooling) and heat removal capability (e.g.,
suppression pool cooling) are unavailable in boiling water reactors
(BWRs), reactor depressurization in conjunction with low pressure
makeup (e.g., low pressure coolant injection) and heat removal
capability (e.g., shutdown cooling) can be used to cool the core. A 10%
failure probability for recovery actions to provide core cooling using
alternative means is assumed for Diablo Canyon, the only West Coast PWR
plant
[[Page 68416]]
with F&B capability, when a snubber impacting more than one train of
the AFW system (i.e., when LCO 3.0.8b is applicable) is out of service.
This failure probability value is significantly higher than the value
of 2.2E-2 used in Diablo Canyon's PRA. Furthermore, Diablo Canyon has
analyzed the impact of a single limiting snubber failure, and concluded
that no single snubber failure would impact two trains of AFW. No
credit for recovery actions to provide core cooling using alternative
means is necessary for West Coast PWR plants with no F&B capability
because it has been determined that there is no single snubber whose
non-functionality would disable two trains of AFW in a seismic event of
magnitude up to the plant's safe shutdown earthquake (SSE). It should
be noted that a similar credit could have been applied to most Central
and Eastern U.S. plants but this was not necessary to demonstrate the
low risk impact of the proposed TS change due to the lower earthquake
frequencies at Central and Eastern U.S. plants as compared to West
Coast plants.
The earthquake frequency at the 0.1g level was assumed to
be 1E-3/year for Central and Eastern U.S. plants and 1E-1/year for West
Coast plants. Each of these two values envelop the range of earthquake
frequency values at the 0.1g level, for Eastern U.S. and West Coast
sites, respectively (References 5 and 7).
The risk impact associated with non-LOOP accident
sequences (e.g., seismically initiated loss-of-coolant-accident (LOCA)
or anticipated-transient-without-scram (ATWS) sequences) was not
assessed. However, this risk impact is small compared to the risk
impact associated with the LOOP accident sequences modeled in the
simplified bounding risk assessment. Non-LOOP accident sequences, due
to the ruggedness of nuclear power plant designs, require seismically-
induced failures that occur at earthquake levels above 0.3g. Thus, the
frequency of earthquakes initiating non-LOOP accident sequences is much
smaller than the frequency of seismically-initiated LOOP events.
Furthermore, because of the conservative assumption made for LOOP
sequences that a 0.1g level earthquake would fail all piping associated
with inoperable snubbers, non-LOOP sequences would not include any more
failures associated with inoperable snubbers than LOOP sequences.
Therefore, the risk impact of inoperable snubbers associated with non-
LOOP accident sequences is small compared to the risk impact associated
with the LOOP accident sequences modeled in the simplified bounding
risk assessment.
The risk impact of dynamic loadings other than seismic
loads is not assessed. These shock-type loads include thrust loads,
blowdown loads, waterhammer loads, steamhammer loads, LOCA loads and
pipe rupture loads. However, there are some important distinctions
between non-seismic (shock-type) loads and seismic loads which indicate
that, in general, the risk impact of the out-of-service snubbers is
smaller for non-seismic loads than for seismic loads. First, while a
seismic load affects the entire plant, the impact of a non-seismic load
is localized to a certain system or area of the plant. Second, although
non-seismic shock loads may be higher in total force and the impact
could be as much or more than seismic loads, generally they are of much
shorter duration than seismic loads. Third, the impact of non-seismic
loads is more plant specific, and thus harder to analyze generically,
than for seismic loads. For these reasons, licensees will be required
to perform an engineering assessment every time LCO 3.0.8 is used and
show that at least one train of each system that is supported by the
inoperable snubber(s) would remain capable of performing their required
safety or support functions for postulated design loads other than
seismic loads.
3.1 Risk Assessment Results and Insights
The results and insights from the implementation of the three-
tiered approach of RG 1.177 to support the proposed addition of LCO
3.0.8 to the TS are summarized and evaluated in the following sections
3.1.1 to 3.1.3.
3.1.1 Risk Impact
The bounding risk assessment approach, discussed in section 3.0,
was implemented generically for all U.S. operating nuclear power
plants. Risk assessments were performed for two categories of plants,
Central and East Coast plants and West Coast plants, based on
historical seismic hazard curves (earthquake frequencies and associated
magnitudes). The first category, Central and East Coast plants,
includes the vast majority of the U.S. nuclear power plant population
(Reference 7). For each category of plants, two risk assessments were
performed:
The first risk assessment applies to cases where all
inoperable snubbers are associated with only one train (or subsystem)
of the impacted safety systems. It was conservatively assumed that a
single train (or subsystem) of each safety system is unavailable. It
was also assumed that the probability of non-mitigation using the
unaffected redundant trains (or subsystems) is 2%. This is a
conservative value given that for core damage to occur under those
conditions, two or more failures are required.
The second risk assessment applies to the case where one
or more of the inoperable snubbers are associated with multiple trains
(or subsystems) of the same safety systems. It was assumed in this
bounding analysis that all safety systems are unavailable to mitigate
the accident, except for West Coast PWR plants. Credit for using F&B to
provide core cooling is taken for plants having F&B capability (e.g.,
Diablo Canyon) when a snubber impacting more than one train of the AFW
system is inoperable. Credit for one AFW train to provide core cooling
is taken for West Coast PWR plants with no F&B capability (e.g., San
Onofre) because it has been determined that there is no single snubber
whose non-functionality would disable two trains of AFW in a seismic
event of magnitude up to the plant's safe shutdown earthquake (SSE).
The results of the performed risk assessments, in terms of core
damage and large early release risk impacts, are summarized in Table 1.
The first row lists the conditional risk increase, in terms of CDF
(core damage frequency), [Delta]RCDF, caused by the out-of-
service snubbers (as assumed in the bounding analysis). The second and
third rows list the ICCDP (incremental conditional core damage
probability) and the ICLERP (incremental conditional large early
release probability) values, respectively. The ICCDP for the case where
all inoperable snubbers are associated with only one train (or
subsystem) of the supported safety systems, was obtained by multiplying
the corresponding [Delta]RCDF value by the time fraction of
the proposed 72-hour delay to enter the actions for the supported
equipment. The ICCDP for the case where one or more of the inoperable
snubbers are associated with multiple trains (or subsystems) of the
same safety system, was obtained by multiplying the corresponding
[Delta]RCDF value by the time fraction of the proposed 12-
hour delay to enter the actions for the supported equipment. The ICLERP
values were obtained by multiplying the corresponding ICCDP values by
0.1 (i.e., by assuming that the ICLERP value is an order of magnitude
less than the ICCDP). This assumption is conservative since containment
bypass scenarios, such as steam generator tube rupture accidents and
interfacing system loss-of-coolant accidents, would not be
[[Page 68417]]
uniquely affected by the out-of-service snubbers. Finally, the fourth
and fifth rows list the assessed [Delta]CDF and [Delta]LERF values,
respectively. These values were obtained by dividing the corresponding
ICCDP and ICLERP values by 1.5 (i.e., by assuming that the snubbers are
tested every 18 months, as was the case before the snubbers were
relocated to a licensee-controlled document). This assumption is
reasonable because (1) it is not expected that licensees would test the
snubbers more often than what used to be required by the TS, and (2)
testing of snubbers is associated with higher risk impact than the
average corrective maintenance of snubbers found inoperable by visual
inspection (testing is expected to involve significantly more snubbers
out of service than corrective maintenance). The assessed [Delta]CDF
and [Delta]LERF values are compared to acceptance guidelines,
consistent with the Commission's Safety Goal Policy Statement as
documented in RG 1.174, so that the plant's average baseline risk is
maintained within a minimal range. This comparison indicates that the
addition of LCO 3.0.8 to the existing TS would have an insignificant
risk impact.
Table 1.--Bounding Risk Assessment Results for Snubbers Impacting a Single Train and Multiple Trains of a
Supported System
----------------------------------------------------------------------------------------------------------------
Central and east coast plants West coast plants
-----------------------------------------------------------------------------
Single train Multiple train Single train Multiple train
----------------------------------------------------------------------------------------------------------------
[Delta]RCDF/yr.................... 1E-6 5E-6 1E-4 5E-4
ICCDP............................. 8E-9 7E-9 8E-7 7E-7
ICLERP............................ 8E-10 7E-10 8E-8 7E-8
[Delta]CDF/yr..................... 5E-9 5E-9 5E-7 5E-7
[Delta]LERF/yr.................... 5E-10 5E-10 5E-8 5E-8
----------------------------------------------------------------------------------------------------------------
The assessed [Delta]CDF and [Delta]LERF values meet the acceptance
criteria of 1E-6/year and 1E-7/year, respectively, based on guidance
provided in RG 1.174. This conclusion is true without taking any credit
for the removal of potential undesirable consequences associated with
the current inconsistent treatment of snubbers (e.g., reduced snubber
testing frequency, increased safety system unavailability and treatment
of snubbers impacting multiple trains) discussed in Section 1 above,
and given the bounding nature of the risk assessment.
The assessed ICCDP and ICLERP values are compared to acceptance
guidelines provided in RG 1.177, which aim at ensuring that the plant
risk does not increase unacceptably during the period the equipment is
taken out of service. This comparison indicates that the addition of
LCO 3.0.8 to the existing TS meets the RG 1.177 numerical guidelines of
5E-7 for ICCDP and 5E-8 for ICLERP. The small deviations shown for West
Coast plants are acceptable because of the bounding nature of the risk
assessments, as discussed in section 2.
The risk assessment results of Table 1 are also compared to
guidance provided in the revised section 11 of NUMARC 93-01, Revision 2
(Reference 8), endorsed by RG 1.182 (Reference 9), for implementing the
requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65.
Such guidance is summarized in Table 2. Guidance regarding the
acceptability of conditional risk increase in terms of CDF (i.e.,
[Delta]RCDF) for a planned configuration is provided. This
guidance states that a specific configuration that is associated with a
CDF higher than 1E-3/year should not be entered voluntarily. Since the
assessed conditional risk increase, [Delta]RCDF, is
significantly less than 1E-3/year, plant configurations including out
of service snubbers and other equipment may be entered voluntarily if
supported by the results of the risk assessment required by 10 CFR
50.65(a)(4), by LCO 3.0.8, or by other TS.
Table 2.--Guidance for Implementing 10 CFR 50.65(a)(4)
------------------------------------------------------------------------
[Delta]RCDF Guidance
------------------------------------------------------------------------
Greater than 1E-3/year................. Configuration should not
normally be entered
voluntarily.
------------------------------------------------------------------------
ICCDP Guidance ICLERP
------------------------------------------------------------------------
Greater than 1E-5............. Configuration should Greater than 1E-
not normally be 6.
entered voluntarily.
1E-6 to 1E-5.................. Assess non- 1E-7 to 1E-6.
quantifiable factors.
Establish risk
management actions..
Less than 1E-6................ Normal work controls.. Less than 1E-7.
------------------------------------------------------------------------
Guidance regarding the acceptability of ICCDP and ICLERP values for
a specific planned configuration and the establishment of risk
management actions is also provided in NUMARC 93-01. This guidance, as
shown in Table 2, states that a specific plant configuration that is
associated with ICCDP and ICLERP values below 1E-6 and 1E-7,
respectively, is considered to require ``normal work controls.'' Table
1 shows that for the majority of plants (i.e., for all plants in the
Central and East Coast category) the conservatively assessed ICCDP and
ICLERP values are over an order of magnitude less than what is
recommended as the threshold for the ``normal work controls'' region.
For West Coast plants, the conservatively assessed ICCDP and ICLERP
values are still within the ``normal work controls'' region. Thus, the
risk contribution from out of service snubbers is within the normal
range of maintenance activities carried out at a plant. Therefore,
plant configurations involving out of service snubbers and other
equipment may be entered voluntarily if supported by the results of the
risk assessment required by 10 CFR
[[Page 68418]]
50.65(a)(4), by LCO 3.0.8, or by other TS. However, this simplified
bounding analysis indicates that for West Coast plants the provisions
of LCO 3.0.8 must be used cautiously and in conjunction with
appropriate management actions, especially when equipment other than
snubbers is also inoperable, based on the results of configuration-
specific risk assessments required by 10 CFR 50.65(a)(4), by LCO 3.0.8,
or by other TS.
The staff finds that the risk assessment results support the
proposed addition of LCO 3.0.8 to the TS. The risk increases associated
with this TS change will be insignificant based on guidance provided in
RGs 1.174 and 1.177 and within the range of risks associated with
normal maintenance activities. In addition, LCO 3.0.8 will remove
potential undesirable consequences stemming from the current
inconsistent treatment of snubbers in the TS, such as reduced frequency
of snubber testing, increased safety system unavailability and the
treatment of snubbers impacting multiple trains.
3.1.2 Identification of High-Risk Configurations
The second tier of the three-tiered approach recommended in RG
1.177 involves the identification of potentially high-risk
configurations that could exist if equipment, in addition to that
associated with the TS change, were to be taken out of service
simultaneously. Insights from the risk assessments, in conjunction with
important assumptions made in the analysis and defense-in-depth
considerations, were used to identify such configurations. To avoid
these potentially high-risk configurations, specific restrictions to
the implementation of the proposed TS changes were identified.
For cases where all inoperable snubbers are associated with only
one train (or subsystem) of the impacted systems (i.e., when LCO 3.0.8a
applies), it was assumed in the analysis that there will be unaffected
redundant trains (or subsystems) available to mitigate the seismically
initiated LOOP accident sequences. This assumption implies that there
will be at least one success path available when LCO 3.0.8a applies.
Therefore, potentially high-risk configurations can be avoided by
ensuring that such a success path exists when LCO 3.0.8a applies. Based
on a review of the accident sequences that contribute to the risk
increase associated with LCO 3.0.8a, as modeled by the simplified
bounding analysis (i.e., accident sequences initiated by a seismically-
induced LOOP event with concurrent loss of all safety system trains
supported by the out of service snubbers), the following restrictions
were identified to prevent potentially high-risk configurations:
For PWR plants, at least one AFW train (including a
minimum set of supporting equipment required for its successful
operation) not associated with the inoperable snubber(s), must be
available when LCO 3.0.8a is used
For BWR plants, one of the following two means of heat
removal must be available when LCO 3.0.8a is used:
--At least one high pressure makeup path (e.g., using high pressure
coolant injection (HPCI) or reactor core isolation cooling (RCIC) or
equivalent) and heat removal capability (e.g., suppression pool
cooling), including a minimum set of supporting equipment required for
success, not associated with the inoperable snubber(s), or
--At least one low pressure makeup path (e.g., low pressure coolant
injection (LPCI) or containment spray (CS)) and heat removal capability
(e.g., suppression pool cooling or shutdown cooling), including a
minimum set of supporting equipment required for success, not
associated with the inoperable snubber(s).
For cases where one or more of the inoperable snubbers are
associated with multiple trains (or subsystems) of the same safety
system (i.e., when LCO 3.0.8b applies), it was assumed in the bounding
analysis that all safety systems are unavailable to mitigate the
accident, except for West Coast plants. Credit for using F&B to provide
core cooling is taken for plants having F&B capability (e.g., Diablo
Canyon) when a snubber impacting more than one train of the AFW system
is inoperable. Credit for one AFW train to provide core cooling is
taken for West Coast PWR plants with no F&B capability (e.g., San
Onofre) because it has been determined that there is no single snubber
whose non-functionality would disable more than one train of AFW in a
seismic event of magnitude up to the plant's safe shutdown earthquake
(SSE). Based on a review of the accident sequences that contribute to
the risk increase associated with LCO 3.0.8b (as modeled by the
simplified bounding analysis) and defense-in-depth considerations, the
following restrictions were identified to prevent potentially high-risk
configurations:
LCO 3.0.8b cannot be used at West Coast PWR plants with no
F&B capability when a snubber whose non-functionality would disable
more than one train of AFW in a seismic event of magnitude up to the
plant's safe shutdown earthquake (SSE) is inoperable (it should be
noted, however, that based on information provided by the industry,
there is no plant that falls in this category).
When LCO 3.0.8b is used at PWR plants, at least one AFW
train (including a minimum set of supporting equipment required for its
successful operation) not associated with the inoperable snubber(s), or
some alternative means of core cooling (e.g., F&B, fire water system or
``aggressive secondary cooldown'' using the steam generators) must be
available.
When LCO 3.0.8b is used at BWR plants, it must be verified
that at least one success path exists, using equipment not associated
with the inoperable snubber(s), to provide makeup and core cooling
needed to mitigate LOOP accident sequences.
3.1.3 Configuration Risk Management
The third tier of the three-tiered approach recommended in RG 1.177
involves the establishment of an overall configuration risk management
program (CRMP) to ensure that potentially risk-significant
configurations resulting from maintenance and other operational
activities are identified. The objective of the CRMP is to manage
configuration-specific risk by appropriate scheduling of plant
activities and/or appropriate compensatory measures. This objective is
met by licensee programs to comply with the requirements of paragraph
(a)(4) of the Maintenance Rule (10 CFR 50.65) to assess and manage risk
resulting from maintenance activities, and by the TS requiring risk
assessments and management using (a)(4) processes if no maintenance is
in progress. These programs can support licensee decision making
regarding the appropriate actions to manage risk whenever a risk-
informed TS is entered. Since the 10 CFR 50.65(a)(4) guidance, section
11 of NUMARC 93-01, does not currently address seismic risk,
implementation guidance must be developed by licensees adopting this
change to ensure that the proposed LCO 3.0.8 is considered with respect
to other plant maintenance activities and integrated into the existing
10 CFR 50.65(a)(4) process whether the process is invoked by a TS or
(a)(4) itself.
3.2 Summary and Conclusions
The option to relocate the snubbers to a licensee controlled
document, as part of the conversion to Improved STS, has resulted in
non-uniform and inconsistent treatment of snubbers. Some potential
undesirable
[[Page 68419]]
consequences of this inconsistent treatment of snubbers are:
Performance of testing during crowded windows when the
supported system is inoperable with the potential to reduce the snubber
testing to a minimum since the relocated snubber requirements are
controlled by the licensee.
Performance of testing during crowded windows when the
supported system is inoperable with the potential to increase the
unavailability of safety systems.
Performance of testing and maintenance on snubbers
affecting multiple trains of the same supported system during the 7
hours allotted before entering MODE 3 under limiting condition of
operation (LCO) 3.0.3.
To remove the inconsistency among plants in the treatment of
snubbers, licensees are proposing a risk-informed TS change which
introduces a delay time before entering the actions for the supported
equipment when one or more snubbers are found inoperable or removed for
testing. Such a delay time will provide needed flexibility in the
performance of maintenance and testing during power operation and at
the same time will enhance overall plant safety by (1) avoiding
unnecessary unscheduled plant shutdowns, thus, minimizing plant
transition and realignment risks; (2) avoiding reduced snubber testing,
thus, increasing the availability of snubbers to perform their
supporting function; (3) performing most of the required testing and
maintenance during the delay time when the supported system is
available to mitigate most challenges, thus, avoiding increases in
safety system unavailability; and (4) providing explicit risk-informed
guidance in areas in which that guidance currently does not exist, such
as the treatment of snubbers impacting more than one redundant train of
a supported system.
The risk impact of the proposed TS changes was assessed following
the three-tiered approach recommended in RG 1.177. A simplified
bounding risk assessment was performed to justify the proposed TS
changes. This bounding assessment assumes that the risk increase
associated with the proposed addition of LCO 3.0.8 to the TS is
associated with accident sequences initiated by a seismically-induced
LOOP event with concurrent loss of all safety system trains supported
by the out of service snubbers. In the case of snubbers associated with
more than one train, it is assumed that all affected trains of the
supported system are failed. This assumption was introduced to allow
the performance of a simple bounding risk assessment approach with
application to all plants and was selected due to the lack of detailed
plant-specific seismic risk assessments for most plants and the lack of
fragility data for piping when one or more supporting snubbers are
inoperable. The impact from the addition of the proposed LCO 3.0.8 to
the TS on defense-in-depth was also evaluated in conjunction with the
risk assessment results.
Based on this integrated evaluation, the staff concludes that the
proposed addition of LCO 3.0.8 to the TS would lead to insignificant
risk increases, if any. Indeed, this conclusion is true without taking
any credit for the removal of potential undesirable consequences
associated with the current inconsistent treatment of snubbers, such as
the effects of avoiding a potential reduction in the snubber testing
frequency and increased safety system unavailability. To be consistent
with the staff's approval, licensees interested in implementing LCO
3.0.8 must, as applicable, operate in accordance with the following
stipulations:
1. Appropriate plant procedures and administrative controls will be
used to implement the following Tier 2 Restrictions.
(a) At least one AFW train (including a minimum set of supporting
equipment required for its successful operation) not associated with
the inoperable snubber(s), must be available when LCO 3.0.8a is used at
PWR plants.
(b) At least one AFW train (including a minimum set of supporting
equipment required for its successful operation) not associated with
the inoperable snubber(s), or some alternative means of core cooling
(e.g., F&B, fire water system or ``aggressive secondary cooldown''
using the steam generators) must be available when LCO 3.0.8b is used
at PWR plants.
(c) LCO 3.0.8b cannot be used by West Coast PWR plants with no F&B
capability when a snubber, whose non-functionality would disable more
than one train of AFW in a seismic event of magnitude up to the plant's
safe shutdown earthquake (SSE), is inoperable.
(d) BWR plants must verify, every time the provisions of LCO 3.0.8
are used, that at least one success path, involving equipment not
associated with the inoperable snubber(s), exists to provide makeup and
core cooling.
(e) Every time the provisions of LCO 3.0.8 are used licensees will
be required to perform a risk assessment, and an operability assessment
to show that at least one train (or subsystem) of systems supported by
the inoperable snubbers would remain capable of performing their
required safety or support functions for postulated design loads other
than seismic loads. The operability assessment, consistent with the
plants licensing design basis, must be documented and available for
inspection by the staff.
2. Should licensees implement the provisions of LCO 3.0.8 for
snubbers, which include delay times to enter the actions for the
supported equipment when one or more snubbers are out of service for
maintenance or testing, it must be done in accordance with an overall
configuration risk management program (CRMP) to ensure that potentially
risk-significant configurations resulting from maintenance and other
operational activities are identified and avoided, as discussed in the
proposed TS Bases. This objective is met by licensee programs to comply
with the requirements of paragraph (a)(4) of the Maintenance Rule, 10
CFR 50.65, to assess and manage risk resulting from maintenance
activities or when this process is invoked by LCO 3.0.8 or other TS.
These programs can support licensee decision making regarding the
appropriate actions to manage risk whenever a risk-informed TS is
entered. Since the 10 CFR 50.65 (a)(4) guidance, Section 11 of NUMARC
93-01, does not currently address seismic risk, implementation guidance
must be developed by licensees adopting this change to ensure that the
proposed LCO 3.0.8 is considered in conjunction with other plant
maintenance activities and integrated into the existing 10 CFR 50.65
(a)(4) process.
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. [For licensees adding a Bases Control Program: The
amendment also changes record keeping, reporting, or administrative
procedures or 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
[[Page 68420]]
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. TSTF-372, Revision 4, ``Addition of LCO 3.0.8, Inoperability of
Snubbers,'' April 23, 2004.
2. Regulatory Guide 1.174, ``An Approach for Using Probabilistic
Risk Assessment in Risk-Informed Decision Making on Plant Specific
Changes to the Licensing Basis,'' USNRC, August 1998.
3. Regulatory Guide 1.177, ``An Approach for Plant Specific Risk-
Informed Decision Making: Technical Specifications,'' USNRC, August
1998.
4. Budnitz, R. J. et al., ``An Approach to the Quantification of
Seismic Margins in Nuclear Power Plants,'' NUREG/CR-4334, Lawrence
Livermore National Laboratory, July 1985.
5. Advanced Light Water Reactor Utility Requirements Document,
Volume 2, ALWR Evolutionary Plant, PRA Key Assumptions and
Groundrules, Electric Power Research Institute, August 1990.
6. Bier V. M. et al., ``Development and Application of a
Comprehensive Framework for Assessing Alternative Approaches to
Snubber Reduction,'' International Topical Conference on
Probabilistic Safety Assessment and Risk Management PSA '87, Swiss
Federal Institute of Technology, Zurich, August 30-September 4,
1987.
7. NUREG-1488, ``Revised Livermore Seismic Hazard Estimates for
Sixty-Nine Nuclear Power Plant Sites East of the Rocky Mountains,''
April 1994.
8. NEI, Revised Section 11 of Revision 2 of NUMARC 93-01, May 2000.
9. Regulatory Guide 1.182, ``Assessing and Managing Risk Before
Maintenance Activities at Nuclear Power Plants,'' May 2000.
Proposed No-Significant-Hazards-Consideration Determination
Description of Amendment Request: A change is proposed to the
standard technical specifications (STS)(NUREGs 1430 through 1434) and
plant specific technical specifications (TS), to allow a delay time for
entering a supported system technical specification (TS) when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of 10 CFR 50.65(a)(4). LCO 3.0.8 will be added to
individual TS providing this allowance.
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 allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges. Therefore,
the probability of an accident previously evaluated is not
significantly increased, if at all. The consequences of an accident
while relying on allowance provided by proposed LCO 3.0.8 are no
different than the consequences of an accident while relying on the TS
required actions in effect without the allowance provided by proposed
LCO 3.0.8. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The addition
of a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. 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).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is assessed
and managed, 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
evaluated. The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Thus, this change does not create the possibility of a new or different
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic event
requiring snubbers is a low-probability occurrence and the overall TS
system safety function would still be available for the vast majority
of anticipated challenges. The risk impact of the proposed TS changes
was assessed following the three-tiered approach recommended in RG
1.177. A bounding risk assessment was performed to justify the proposed
TS changes. This application of LCO 3.0.8 is predicated upon the
licensee's performance of a risk assessment and the management of plant
risk. The net change to the margin of safety is insignificant.
Therefore, this change does 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 no-significant-hazards consideration.
Dated at Rockville, Maryland, this 18th day of November, 2004.
For the Nuclear Regulatory Commission.
Thomas H. Boyce,
Section Chief, Technical Specifications Section, Operating Improvements
Branch, Division of Inspection Program Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-26008 Filed 11-23-04; 8:45 am]
BILLING CODE 7590-01-P