Notice of Availability of Model Application Concerning Technical Specification Improvement To Modify Requirements Regarding the Addition of Limiting Condition for Operation 3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item Improvement Process, 23252-23262 [E5-2171]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 85 (Wednesday, May 4, 2005)]
[Notices]
[Pages 23252-23262]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E5-2171]


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NUCLEAR REGULATORY COMMISSION


Notice of Availability of Model Application Concerning Technical 
Specification Improvement To Modify Requirements Regarding the Addition 
of Limiting Condition for Operation 3.0.8 on the Inoperability of 
Snubbers Using the Consolidated Line Item Improvement Process

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of availability.

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SUMMARY: Notice is hereby given that the staff of the Nuclear 
Regulatory Commission (NRC) has prepared a model application relating 
to the modification of requirements regarding the impact of inoperable 
snubbers not in technical specifications, on supported systems in 
technical specifications (TS). The purpose of this model is to permit 
the NRC to efficiently process amendments that propose to modify 
requirements by adding to the TS a limiting condition for operation 
(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, as generically approved by this notice. 
Licensees of nuclear power reactors to which the model applies could 
request amendments utilizing the model application.

DATES: The NRC staff issued a Federal Register Notice (69 FR 68412, 
November 24, 2004) which provided a Model Safety Evaluation (SE) 
relating to modification of requirements regarding the addition \1\ to 
the TS of LCO 3.0.8 on the impact of inoperable snubbers; similarly the 
NRC staff herein provides a Model Application, including a revised 
Model Safety Evaluation. The NRC staff can most efficiently consider 
applications based upon the Model Application, which references the 
Model Safety Evaluation, if the application is submitted within one 
year of this Federal Register notice.
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    \1\ In conjunction with the proposed change, technical 
specification (TS) requirements for a Bases Control Program, 
consistent with the TS-Bases Control Program described in section 
5.5 of the applicable vendor's standard TS (STS), shall be 
incorporated into the licensee's TS, if not already in the TS.

FOR FURTHER INFORMATION CONTACT: Tom Boyce, Mail Stop: O-12H2, Division 
of Inspection Program Management, Office of Nuclear Reactor Regulation, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
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telephone 301-415-0184.

SUPPLEMENTARY INFORMATION:

Background

    Regulatory Issue Summary 2000-06, ``Consolidated Line Item 
Improvement Process for Adopting Standard Technical Specifications 
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. This is accomplished 
by processing proposed changes to the standard technical specifications 
(STS) in a manner that supports subsequent license amendment 
applications. The CLIIP includes an opportunity for the public to 
comment on proposed changes to the STS following a preliminary 
assessment by the NRC staff and finding that the change will likely be 
offered for adoption by licensees. The CLIIP directs the NRC staff to 
evaluate any comments received for a proposed change to the STS and to 
either reconsider the change or to proceed with announcing the 
availability of the change for proposed adoption by licensees. Those 
licensees opting to apply for the subject change to technical 
specifications are responsible for reviewing the staff's evaluation, 
referencing the applicable technical justifications, and providing any 
necessary plant-specific information. Each amendment application made 
in response to the notice of availability will be processed and noticed 
in accordance with applicable rules and NRC procedures.
    This notice involves the modification of requirements regarding the 
addition to the TS of 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. This change was 
proposed for incorporation into the standard technical specifications 
by all Owners Groups participants in the Technical Specification Task 
Force (TSTF) and is designated TSTF-372 Revision 4, which was 
referenced in the Federal Register Notice (FRN) 69 FR 68412, of 
November 24, 2004, and can both be viewed on the NRC's Web page at 
https://www.nrc.gov/reactors/operating/licensing/techspecs.html.

Applicability

    This proposed change to modify technical specification requirements 
for the impact of inoperable non-technical specification snubbers on 
supported systems in TS is applicable to all licensees who currently 
have or who will adopt, in conjunction with the proposed change, 
technical specification requirements for a Bases control program 
consistent with the

[[Page 23253]]

Technical Specifications 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 each licensee applying for the changes addressed by 
TSTF-372 Revision 4 using the CLIIP to include the Bases for the 
proposed technical specifications. In addition, for those licensees 
that have not adopted requirements for a Bases control program by 
converting to the improved STS or by other means, the staff requests 
that you include the requirements for a Bases control program 
consistent with the STS in your request 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 surveillance requirement (SR) 
3.0.8. The staff is requesting that the Bases be included with the 
proposed license amendments because, in this case, the changes to the 
technical specifications and changes to the associated Bases form an 
integrated change to a plant's licensing bases. To ensure that the 
overall change, including the Bases, includes the appropriate 
regulatory controls, the staff plans to condition the issuance of each 
license amendment on incorporation of the changes to the Bases document 
and on ensuring the licensee's TS have a Bases Control Program for 
controlling changes to the Bases. The CLIIP does not prevent licensees 
from requesting an alternative approach or proposing the changes 
without the requested Bases and Bases control program. Variations from 
the approach recommended in this notice may, however, require 
additional justification, additional review by the NRC staff and may 
increase the time and resources needed for the review.

Public Notices

    The staff issued a Federal Register Notice (69 FR 68412, November 
24, 2004) that requested public comment on the NRC's pending action to 
approve modification of TS requirements regarding the impact of 
inoperable non-technical specification snubbers on supported systems in 
TS. In particular, following an assessment and draft safety evaluation 
by the NRC staff, the staff sought public comment on proposed changes 
to the STS, designated as TSTF-372 Revision 4. The TSTF-372 Revision 4 
can be viewed on the NRC's Web page at https://www.nrc.gov/reactors/
operating/licensing/techspecs.html. TSTF-372 Revision 4 may be 
examined, and/or copied for a fee, at the NRC's Public Document Room, 
located at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records are accessible 
electronically from the ADAMS Public Library component on the NRC Web 
site (the Electronic Reading Room), at https://www.nrc.gov/reading-rm/
adams.html.
    In response to the notice soliciting comments from interested 
members of the public about modifying the TS requirements regarding the 
impact of inoperable non-technical specification snubbers on supported 
systems in TS, the staff received three sets of comments (from 
licensees and the TSTF Owners Groups, representing licensees). Specific 
comments on the model SE were offered, and are summarized and discussed 
below:
    1. Comment: Performing and documenting the engineering assessment 
every time LCO 3.0.8 is used is unnecessary as it is unlikely that the 
design function of the snubbers will change. The Safety Evaluation 
should be revised to state that when LCO 3.0.8 is used, licensees must 
confirm that at least one train of each system that is supported by the 
inoperable snubber(s) would remain capable of performing its required 
safety or support functions for postulated design loads other than 
seismic loads.
    The evaluation described is not an ``operability assessment.'' In 
order for LCO 3.0.8 to be needed, the system supported by the snubber 
to be removed from service would not be considered operable. The 
phrases ``operability assessment'' and ``engineering assessment'' 
should be replaced as described in the previous bullet.
    Response: The terms ``engineering assessment'' and ``operability 
assessment'' were used to describe the determination licensees must 
make, when a snubber is inoperable, that the snubber is seismic or non-
seismic in function, the number of trains affected, and that the 
underlying assumptions of LCO 3.0.8 apply, before invoking LCO 3.0.8. 
It is recognized that the determination is only required when the 
inoperable snubber is required to support a system that is required to 
be operable by a TS, and when that TS is in a mode of applicability. 
Also, when a train is removed from service for maintenance, the risk 
assessment for the performance of the maintenance would encompass that 
for snubbers supporting only equipment on that train. So there are 
circumstances in which assessments/determinations for inoperable 
snubbers are not required. In recognition of the variability of the 
degree of determination required for an inoperable snubber, and the 
fact that the term ``assessment'' has formal procedural connotations, 
the wording has been changed as suggested, to require that `` * * * 
licensees confirm * * * '' and not assess, every time a snubber is 
inoperable.
    2. Comment: In [section 3.2] item 1.(e), the Safety Evaluation uses 
the phrase ``perform a risk assessment.'' This phrase also appears on 
page 68420 of the Federal Register notice, third column, in the No 
Significant Hazards Consideration (NSHC), Criterion 3 discussion. The 
proposed Technical Specifications state that ``risk must be assessed 
and managed.'' Item 1.(e) and the NSHC should be revised to be 
consistent with the proposed Technical Specifications.
    Response: The staff agrees. The wording will be changed to be 
consistent with 10 CFR 50.65(a)(4), which requires the licensee to 
``assess and manage the increase in risk.''
    3. Comment: Documenting the design functions of the snubber(s) for 
NRC inspection should not be required. As stated in TSTF-372, the risk 
assessments will be consistent with those performed to meet the 
requirements of 10 CFR 50.65(a)(4). It is not required that the risk 
assessments performed to meet the requirements of 10 CFR 50.65(a)(4) be 
documented. It would be inconsistent to require documentation of the 
particular portion of the 10 CFR 50.65(a)(4) risk assessments related 
to snubbers. In addition, this information exists in the plant's design 
documentation and it imposes an unnecessary burden on the licensee to 
record for this particular purpose otherwise generic information.
    Response: To be consistent with the requirements of 10 CFR 
50.65(a)(4), which does not require the documentation discussed in this 
comment, and in light of the variability of assessments associated with 
inoperable snubbers (as noted in the response to comment 1 above), the 
requirement for every evaluation to be documented has been removed. The 
staff nonetheless considers that it would be prudent in many 
circumstances for the evaluation to be documented, and that it would 
also be efficient if licensees were able to refer to prior evaluations. 
LCO 3.0.8 does not apply to non-seismic snubbers. In addition, a record 
of the design function of the inoperable snubber (i.e., seismic vs. 
non-seismic), implementation of any applicable Tier 2 restrictions, and 
the associated plant configuration shall be available on a recoverable 
basis for staff inspection.
    4. Comment: On page 68415 of the Federal Register Notice, the third

[[Page 23254]]

column, first paragraph, the following statement is made: ``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.'' The statement ``it is not expected 
to have more than a few snubbers supporting a given safety system out 
for testing at a time'' does not appear in TSTF-372 and is not an 
assumption of the risk assessment that was performed to support the 
Traveler. The Traveler risk assessment assumed that the systems 
affected by removed snubbers are unavailable. Therefore, the number of 
removed snubbers is irrelevant. The statement implies that plants must 
impose some undefined limit (i.e., a ``few'') on the number of snubbers 
that can be simultaneously removed from a given system. Such a 
restriction is unnecessary and confusing. It is recommended that the 
sentence be revised to state, ``Since the licensee controlled testing 
is done on only a small (about 10%) representative sample of the total 
snubber population, typically only a few snubbers supporting a given 
safety system are out for testing at a time.'' This changes the 
sentence from what could be construed as a requirement to a statement 
of fact.
    Response: The staff accepts the use of the phrase, ``typically 
only,'' as a substitute; the staff considers the phrases equivalent.
    5. Comment: On page 68419 of the Federal Register Notice, the third 
column, first paragraph prior to Section 4.0, State Consultation, the 
following statement is made: ``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.''
    A similar statement is made on page 68418 of the Federal Register 
Notice, the third column, the last paragraph of Section 3.1.3. It is 
not necessary to develop independent ``implementation guidance'' 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. We recommend that the sentences be revised to 
state: Since the 10 CFR 50.65(a)(4) guidance, Section 11 of NUMARC 93-
01, does not currently address seismic risk, licensees adopting this 
change must 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.
    Response: The staff accepts the wording change. In this case the 
use of the term ``implementation guidance'' was not intended to convey 
formal industry guidance. Therefore, to avoid confusion using the words 
``must ensure'' is preferable. Wording has been added in the Safety 
Evaluation to ensure that seismic risk assessments used to satisfy the 
10 CFR 50.65(a)(4) process will be based upon either detailed seismic 
probabilistic risk assessment (PRA) based evaluations or bounding risk 
analyses, such as utilized in the assessment included in the Safety 
Evaluation.
    6. Comment: On page 68414 of the Federal Register Notice, middle 
column, first paragraph, it is stated that prior to conversion to 
improved STS, the 72-hour delay time provision that was typically 
included in the snubber technical specification was applicable only to 
snubbers found to be inoperable (i.e., emergent conditions only). This 
characterization is contrary to previous NRC positions (see References 
4 and 5 of TSTF-372, Revision 4). It is a long standing industry 
practice to utilize the 72-hour delay for the removal of snubbers for 
maintenance and testing purposes, not only emergent conditions.
    Response: There remain some differing interpretations on what pre-
improved STS allowed. Regardless of prior practices and what older 
specifications permitted, this change will clarify and make consistent 
practices and understanding of what is permitted. Therefore, statements 
of what pre-improved STS allowed are removed from the text.
    7. Comment: In the first paragraph of the Summary, the term ``non-
technical specifications snubbers'' is used. That term is not defined 
or used elsewhere. In section 1.0, INTRODUCTION, the new LCO 3.0.8 
identifies the snubbers of interest as ``required snubbers.'' In 
section 2.0, Regulatory Evaluation, the snubbers of interest are 
characterized as ``relocated snubbers.''
    Some clarification is requested to ensure that the snubbers of 
interest are clearly understood to be those required to support 
Technical Specifications functions.
    Response: In the first paragraph of the Summary, the term ``non-
technical specifications snubbers'' is changed to ``snubbers not in 
technical specifications.'' In section 1.0, INTRODUCTION, the new LCO 
3.0.8 identifies the snubbers of interest as ``required snubbers.'' In 
technical specifications the term ``required snubbers'' is understood 
to be those required to support Technical Specifications functions. In 
section 2.0, REGULATORY EVALUATION, the term ``relocated snubber 
requirements'' has been changed to ``snubber requirements that have 
been relocated from technical specifications* * *''.
    8. Comment: For licensees who have not converted to the improved 
STS, some clarification is needed for the ``other means'' by which a 
licensee could have adopted a Bases control program. Is it necessary 
that the Bases control program be incorporated into the Technical 
Specifications, or would the establishment of a procedure in the plant 
operating manual be sufficient?
    Response: The Risk Management Technical Specifications (RMTS) 
Initiatives that have been approved to-date have each required the 
adoption of a Bases Control Program, if not previously adopted through 
conversion to the STS. It is necessary that the Bases Control Program 
be incorporated into the TS. At this point it is expected that most 
plants have adopted a Bases Control Program in the Administrative 
Controls Section of their TS. As noted, licensees are not prevented 
from requesting an alternative approach or proposing the changes 
without the requested Bases and Bases control program. Variations from 
the approach recommended in this notice may, however, require 
additional justification, additional review by the NRC staff and may 
increase the time and resources needed for the review. In addition, an 
alternative approach will most likely have to similarly involve a 
change to the plant license.
    9. Comment: Section 3.1.2 of the model safety evaluation regarding 
the use of LCO 3.0.8b for boiling water reactors requires that ``at 
least one success path exists, using equipment not associated with the 
inoperable snubber(s), to provide makeup and cooling needed to mitigate 
LOOP accident sequences.'' The phrase ``needed to mitigate LOOP 
accident sequences'' is absent in the corresponding implementation 
requirements in Section 3.2.1(d), which implies all accident sequences 
must be considered. This phrase should be restored to Section 3.2.1(d) 
to clarify the type of analysis that must be performed.
    Response: The staff agrees. The phrase ``needed to mitigate LOOP 
accident sequences'' is added to Section 3.2.1(d).

    Dated at Rockville, Maryland, this 27th day of April 2005.


[[Page 23255]]


    For the Nuclear Regulatory Commission.
Theodore R. Tjader,
Senior Reactor Engineer, Technical Specifications Section, Operating 
Improvements Branch, Division of Inspection Program Management, Office 
of Nuclear Reactor Regulation.

Model Safety Evaluation

Technical Specification Task Force (TSTF) Change TSTF-372

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.''
    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.

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 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

[[Page 23256]]

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. The industry's proposal would allow 
a time delay for all conditions, including snubber removal for testing 
at power. 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 snubber requirements that 
have been relocated from TS 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.

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 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

[[Page 23257]]

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, typically only 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 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 Cost 
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

[[Page 23258]]

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 confirm every time LCO 3.0.8 is used, 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 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 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

[[Page 23259]]

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 not       Greater than 1E-6.
                                           normally be entered
                                           voluntarily.
1E-6 to 1E-5............................  Assess non-quantifiable        1E-7 to 1E-6.
                                           factors; Establish risk
                                           management actions.
Less than 1E-6..........................  Normal work controls.........  Less than1E-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 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.,

[[Page 23260]]

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 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 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, firewater 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, the 
revised (May 2000) Section 11 of NUMARC 93-01, does not currently 
address seismic risk, licensees adopting this change must 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 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 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 as