68412 Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 

well as the Chairman’s ruling on 

requests for the opportunity to present 

oral statements and the time allotted 

therefor can be obtained by contacting 

Mr. Sam Duraiswamy, Cognizant ACRS 

staff (301–415–7364), between 7:30 a.m. 

and 4:15 p.m., e.t. 

ACRS meeting agenda, meeting 

transcripts, and letter reports are 

available through the NRC Public 

Document Room at pdr@nrc.gov, or by 

calling the PDR at 1–800–397–4209, or 

from the Publicly Available Records 

System (PARS) component of NRC’s 

document system (ADAMS) which is 

accessible from the NRC Web site at 

http://www.nrc.gov/reading-rm/

adams.html or http://www.nrc.gov/

reading-rm/doc-collections/ (ACRS & 

ACNW Mtg schedules/agendas). 

Videoteleconferencing service is 

available for observing open sessions of 

ACRS meetings. Those wishing to use 

this service for observing ACRS 

meetings should contact Mr. Theron 

Brown, ACRS Audio Visual Technician 

(301–415–8066), between 7:30 a.m. and 

3:45 p.m., e.t., at least 10 days before the 

meeting to ensure the availability of this 

service. 

Individuals or organizations 

requesting this service will be 

responsible for telephone line charges 

and for providing the equipment and 

facilities that they use to establish the 

videoteleconferencing link. The 

availability of videoteleconferencing 

services is not guaranteed. 

The ACRS meeting dates for Calendar 

Year 2005 are provided below:

ACRS meeting No. Meeting dates 

............................... January 2005 (No 

meeting). 

519 ............................ February 10–12, 

2005. 

520 ............................ March 3–5, 2005. 

521 ............................ April 7–9, 2005. 

522 ............................ May 5–7, 2005. 

523 ............................ June 1–3, 2005. 

524 ............................ July 6–8, 2005. 

................................ August 2005 (No 

meeting). 

525 ............................ September 8–10, 

2005. 

526 ............................ October 6–8, 2005. 

527 ............................ November 3–5, 2005. 

528 ............................ December 8–10, 

2005. 

Dated: November 18, 2004. 

Andrew L. Bates, 

Advisory Committee Management Officer.

[FR Doc. 04–26005 Filed 11–23–04; 8:45 am] 

BILLING CODE 7590–01–P

NUCLEAR REGULATORY 

COMMISSION 

Notice of Opportunity To Comment on 

Model Safety Evaluation on Technical 

Specification Improvement To 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 

CLIIP@nrc.gov.

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

Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 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 lowprobability 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

68414 Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 

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

Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 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 (DCDF) and the 

average yearly large early release 

frequency (DLERF) and (2) by the 

incremental conditional core damage 

probability (ICCDP) and the incremental 

conditional large early release 

probability (ICLERP). The assessed 

DCDF and DLERF 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 risksignificant 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 risksignificant configurations resulting from 

maintenance and other operational 

activities are identified. The objective of 

the CRMP is to manage configurationspecific 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-offsitepower (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

68416 Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 

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-coolantaccident (LOCA) or anticipatedtransient-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 nonseismic (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 nonseismic 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 nonmitigation 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), 

DRCDF, 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 DRCDF

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 DRCDF 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-ofcoolant accidents, would not be

Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 68417

uniquely affected by the out-of-service 

snubbers. Finally, the fourth and fifth 

rows list the assessed DCDF and DLERF 

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 DCDF and 

DLERF 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 

DRCDF/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 

DCDF/yr ........................................................................................................................................... 5E–9 5E–9 5E–7 5E–7 

DLERF/yr ......................................................................................................................................... 5E–10 5E–10 5E–8 5E–8 

The assessed DCDF and DLERF 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., DRCDF) 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, 

DRCDF, 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) 

DRCDF Guidance 

Greater than 1E–3/year ............................................................................ Configuration should not normally be entered voluntarily. 

ICCDP Guidance ICLERP 

Greater than 1E–5 .......................... Configuration should not normally be entered voluntarily ..................... Greater than 1E–6. 

1E–6 to 1E–5 .................................. Assess non-quantifiable factors .............................................................

Establish risk management actions. 

1E–7 to 1E–6. 

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

68418 Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 

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 nonfunctionality 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 risksignificant configurations resulting from 

maintenance and other operational 

activities are identified. The objective of 

the CRMP is to manage configurationspecific 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

Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 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 nonfunctionality 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

68420 Federal Register / Vol. 69, No. 226 / Wednesday, November 24, 2004 / Notices 

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 nosignificant-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-HazardsConsideration 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 lowprobability 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 lowprobability 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 nosignificant-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] 

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