[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]


-----------------------------------------------------------------------

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.

-----------------------------------------------------------------------

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