ML20198M784

SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2

ML20198M784
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/29/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198M766	List: ML20198M762 ML20198M784
References
FACA, NUDOCS 9901050353
Download: ML20198M784 (22)

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UNITED STATES s

j-NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 30666-0001

.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTO l

PROPOSAL TO USE ASME CODE CASE N-578 AS AN ALTERNA TO ASME CODE SECTION XI. TABLE IWX-2500

~ ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNIT NO. 2 DOCKET NO. 50-368

1.0 INTRODUCTION

Current requirements for conducting inspections at commercial nuclear power plants contained in the 1989 Edition of Section XI, Division 1 of the American Society of Mech Engineers (ASME) Boiler and Pressure Vessel Code, entitled Rules for Inservice l

NuclearPower Plant Components (hereinafter called the Code). In letters dated September 30,1997 ', and March 31,1998 2, the licensee, Entergy Operations, In attemative to the current requirements for the examination of piping welds at Arkansa One, Unit No. 2 (ANO-2). In response to requests for additional infomiation (RAls) fr NRC, the licensee'sent additional information by letters dated October 8,19985, 1998 *, and December 8,1998 5, to address certain items in support of the risk-informed inservice inspection pilot application at ANO-2. The licensee's submittals were reviewe information with NRC guidance documents *7 and applicable portions of the Electric Power Research Institute (EPRI) risk-informed topical report No. TR-106706' as appro iriate be noted that (1) the EPRI report is still under development and is the subject of a s evaluation and (2) this safety evaluation does not constitute approval of the EPR

[

in any manner. Additionally, it should be noted that although the licensee's alternativ the use of ASME Code Case N-578, this evaluation does not provide generic approva endorsement of this Code Case as written. This evaluation applies only to piping syste ANO-2. All Code-required inspections of other non-piping safety-related components shal continue to be performed in accordance with the ASME Section XI inservice inspe I

as required by the licensee's Technical Specifications.

l l

2.0

SUMMARY

OF PROPOSED APPROACH L

The licensee is required to perform inservice inspection (ISI) of ASME Code Categor C-F piping welds during successive 120-month (10-year) intervals. Currently,25% of all Category B-J piping welds greater than 1-inch nominal diameter are selected for volumet surface examination or both on the basis of the existing stress analyses. For Category piping welds,7.5% of non-exempt welds are selected for surface or volumetric examinatio l

both.

I 9901050353 981229 PDR ADOCK 05000368 P

PDR

I-2' j

l Pursuant to '10 CFR 50.55a(3)(i), the licensee has proposed to implement Code Case N-578, 7

Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B, augmented with the more i

detailed provisions of Reference 8, and other methodology enhancements described in this report, as an attemative to the Code requirements for the examination of Class 1 and 2 piping welds at ANO-2. By letter dated September 30,1997, the licensee submitted separate risk evaluations for the following systems:

(1)

High-Pressure Safety injection (2)

Reactor Coolant l

(3)

Chemical and Volume Control (4)

Containment Spray.

(5). Low-Pressure Safety injection and Shutdown Cooling (6)

Emergency Feedwater (7)

Main Feedwater (8)

Main Steam The licensee submitted a formal request for NRC review by letter dated March 31,1998. This letter also contained a risk evaluation for the service water system. In effect, the attemative proposed by the licensee would significantly reduce and re-target inservice examinations, based on the consequences of various piping failure (s) in conjunction with expected degradation i

mechanisms. In response to RAls from the NRC, the licensee sent a letter on October 8,1998, addressing specific items of concem in support of the risk-informed inservice inspection (RI-ISI) pilot application at ANO-2.

1 in accordance with 10 CFR 50.55a(a)(3)(i), the NRC may authorize a proposed alternative to regulatory requirements when the applicant demonstrates that the alternative provides an acceptable level of quality and safety, in this case, the licensee must demonstrate that the proposed alternative provides protection comparable to the requirements of ASME Section XI, which prescribes the how, when, where, and number of examinations to be performed on Class 1 and 2 piping systems. The licensee has requested approval of this attemative for implementation during the January 1999 refueling outage. ANO-2 is currently in its second 10-year ISI interval, which is scheduled to end on March 26< 2000.

The licensee has submitted the proposed alternative on the grounds that Rl-ISI will provide an acceptable level of quality and safety. The licensee stated The principal objective of a risk-informed inservice inspection (ISI) approach is to focus resources on higher risk elements. Entergy believes that achieving this objective will result in increased plant safety and a significant reduction in worker radiation exposure, while also reducing plant operating and maintenance costs.

The licensee's proposed attemative applies specifically to the nondestructive examination (NDE) of Class 1 and 2 piping, but also includes Class 3 systems and certain non-Code-classed piping in the risk evaluations.

The fundamental basis for the proposed a%mative is that the overall plant risk associated with the new process for selection of piping inservice examination locations is essentially the same, I

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. or less than the plant risk resulting from the current ASME Section XI piping ISI selection requirements.

3.0 EVALUATION The licensee's submittals were reviewed with respect to criteria contained in the Standard Review Plan (SRP) Chapter 3.9.8,

• Standard Review Plan For Trial Use for the Review of Risk-Infonned Inservice inspection of Piping, September 1998." The SRP describes the review process and acceptance guidelines for NRC staff reviews of proposed plant-specific, risk-informed changes to a licensee's ISI program for piping. Further guidance in defining acceptable methods for implementing an RI-ISI program is described in Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping,"

which has been issued for trial use and is consistent with the review procedures contained within the SRP. Each section of the SRP, and how the licensee addressed it, is discussed in the sections below.

3.1 Prooosed Chances to ISI Proaram Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee has proposed to implement Code Case N-578 Risk-Informed Requirements for Class 1, 2, and 3 Piping, Mathod B, augmented with the more detailed provisions of EPRI TR-106706, as an alternative to the Code examination requirements for piping systems at the ANO-2 plant. The proposed changes to the ISI program are described in Section 2 of the licensee's submittal. Details of the proposed changes involving the specific pipe systems, segments, numbers of welds selected and revisions to inspection scope, locations, and techniques are given in Section 7 and Appendix C of each submittal. This information was further presented and summarized for all of the RI-ISI systems in the October 8,1998, response to an NRC RAl. In this same response, Entergy also confirmed that the inspection frequency will be the 10-year inspection interval currently required by Section XI, supplemented with augmented inspections for specific degradation mechanisms, i.e., flow accelerated corrosion (FAC), with a frequency as specified in the existing plant program.

Critical plant sa'ety functions and associated success criteria are common to all of the submittals when determining the number of backup trains that are available for the consequence rankings of each pipe segment failure, as identified in Figure 3-1 in conjunction with Table 3-1 of the Service Water Analysis. A more detailed discussion and presentation of each critical safety function was submitted by the licensee in response to an NRC i4Al. The simplified success criteria diagrams were expanded to more clearly depict the relationship between the critical safety functions and the systems that provide these functions In Section 6 of each submittal, plant-specific experience was evaluated (1) to identify any potential degradation mechanisms in the piping pressure boundary associated with particular plant configurations and service conditlons, (2) to supplement the EPRI industry experience review, and (3) to confirm susceptibility to degradation mechanisms. The licensee used a separate independent evaluation'0, in conjunction with the EPRI methodology, to determine applicable degradation mechanisms appropriate for ANO-2. These determinations serve to assess the potential for failure specific and target-specific examination locations within each piping system, in lieu of current Code selection processes.

. In consideration of the information supplied by the licensee in the original submittals and in the responses to NRC RIAs, the licensee has met the SRP requirement to adequately define the proposed changes to the current ISI program with respect to implementation of Code Case N-578 as augmented by EPRI TR-106706, and the methodology enhancements discussed herein.

3.2 Enoineerina Analysis An engineering analysis of the proposed changes is required using a combination of traditional engineering analysis with supporting insights from the probabilistic risk assessment (PRA). The licensee further elaborated on how the engineering analyses conducted for the ANO-2 RI-ISI program ensures that the proposed changes are consistent with the principles of defense-in-depth and that adequate safety margins will be maintained. The licensee does this by evaluating a location's susceptibility to a particular degradation mechanism, which should increase the likelihood of finding flaws or indications that may be precursors to leak or rupture, and then performing an independent assessment of the consequence of a failure in that location.

Further details regarding the engineering analysis and risk-based evaluations are discussed in the next sections.

3.2.1 Traditional Analysis The objectives of ISI and ASME Section XI are to identify conditions (i.e., flaw indications) that are precursors to leaks and ruptures in the pressure boundary that may affect plant safety.

Therefore, the RI-ISI program must meet this objective to be found acceptable for use as an alternative to the requirements in ASME Section XI. The current licensing basis for ISI at ANO-2 includes augmented examinations as a result of NRC-issued notifications. The process for assessing degradation mechanisms includes the consideration of such attributes Es thermal fatigue and stress corrosion cracking in defining a piping segment's susceptibility to these and to other damage processes. For this reason, several of the currently applied augmented examination programs will no longer be separately implemented, but are enveloped by the RI-ISI program. These include augmented examinations as a result of the following:

NRC Bulletin 88-08, " Thermal Stresses in Piping Connected to Reactor Coolant Systems,"

Supplements 1,2, and 3, NRC Bulletin 88-11, " Pressurizer Surge Line Thermal Stratification",

+

NRC Information Notice 93-020, " Thermal Fatigue Cracking of Feedwater Piping to Steam

+

Generators", and IE Bulletin 79-17, " Pipe Cracks in Stagnant Borated Water Systems at PWR Plants."

+

However, the licensee has confirmed that, on the basis of the RI-ISI degradation assessment, the fo!!owing two existing augmented examinations will be maintained at ANO-2:

Flow Accelerated Corrosion (FAC) Inspection Program, and

+

i l

. The classification of "no-break zones", as described in Brr ich Technical Position ME3 3-1," Postulated Rupture Locations in Fluid Systems inside end Outside Containment."

The licensee has also clarified which portions of the Code the proposed ISI changes are intended to replace, and identified the specific changes in moving from the current Section XI program to the RI-ISI program.

The licensee has encompassed appropriate degradation mechanisms within the RI-ISI piping segment evaluations. In addition, certain augmented programs will remain intact on the basis of these determinations. Therefore, the licensee has met the intent of the SRP requirement to confirm that existing augmented examination programs that are part of the current licensing basis for ANO-2, pertaining to the integrity management of piping, will not be adversely impacted by implementation of the proposed attemative.

3.2.2 EfM The ANO-2 Individual Plant Examination (IPE) was completed in August,1992 (ANO-2 PRA 94-R-2005-01, Rev. 0). The IPE estimated a core damage frequency (CDF) of 3.4E-5/yr. A limited scope Level 11 analysis considered all relevant plant damage states, containment systems states, and containment failure modes. Detailed phenomena and structural analysis were not performed. ANO-2 design features were considered in the IPE via limited engineering calculations or on the basis of a reference plant. The licensee defined large early release frequency (LERF) as early containment failure without source term mitigation and reported that this is expected to occur in about 7% of all severe accidents. No LERF was reported, although a frequency of 2.4E-6/yr can be estimated from the findings.

intemal floods were evaluated through a seven-step screening process. Tha licensee estimated a CDF of SE-7/yr from the one zone (of 78 zones) to survive up to the last screening step, but claimed that this estimate was still conservative. This estimate fell below the licensee's 1E-6/yr cut-off criteria for individual sequences, so the licensee reported that all flooding sequences had been screened out.

According to the licensee's October 8,1998, submittal, the licensee updated the PRA between the original Rl-ISI submittals in 1997 and the October 1998 submittal. The updated PRA was used to perform confirmatory calculations to ensure that systems interactions and common cause failures are appropriately reflected in the equivalent number of trains assigned to different systems.

The methodology provides guidance on assigning consequence categories to segment breaks on the basis of the number of available trains, broad categories of initiating event frequencies, and exposure times. As discussed in Section 3.2.6, however, use of the methodology requires confidence that the train unavailabilities, initiating event frequency, and associated conditional core damage probabilities (CCDPs) are consistent with the bounding values assumed by the methodology. The licensee used quantitative PRA results to support the following plant-specific evaluations:

(1)

The !icensee uses PRA results to estimate the CCDP for those segment failures t cause only initiaSng events and that do not cause any loss of mitigating functional capability. The CCDP was obtained by dividing the CDF for each relevant initiat by the initiating event frequency. The CCDP result can be directly compared to guidelines to place the segmer.t in the appropriate consequence category. Th are dicuussed in Section 3.2.6.

(2)

For those segment failures that only fail mitigating systems and do not cau the licensee used the matrix process described in Section 3.2.6. The PRA was used however, to calculate several system and function unavailabilities by setting the events representing the equipment and functions failed by the segment rupture to " fa in the baseline PRA. The results were compared to the 0.01 unaveilability pe defined in the methodology) to ensure that an equivalent number of "back-up train deterministically assigned is consistent with the PRA results.

(3)

For those system failures that fail mitigating system (s) and that cause a plant tr

^

licensee used the ma;rix process dcseribed in Section 3.2.6. The PRA was used, however, to calculate several CCDPs (after setting the equipment and func the segment rupture to failed in the baseline PRA) to ensure that system and equ interactions between the initiating event and the credited, mitigating systems trains w appropriately accounted for.

Recovery actions credited in the baseline PRA but no longer appNable because of damage caused by the segment rupture are factored out of all the CCDPs calculated. T was not directly used to determine any LERF values. LERF considerations were evaluated as described in Section 3.2.6.

The staff finds that the results of the (PE and from the updated PRA were used in a mann consistent with their definitions. The staff also finds that the use of the updated PRA to the requected confirmatory calculations provides additional assurance that the result of the ISI evaluations reflect the current as-operated state of the plant.

Qualitv of PRA Licensee personnel performed the majority of the analytical work during the developme IPE. The IPE program included an independent review team consisting of ANO Operatio Design Engineering Training, and Licensing staff supplemented with contractor enginee The detailed review methodology included the review of initiating events, accident sequences, success criteria, system models and fault trees, human reliability analysis, data evaluation, and containment response evaluation. The licensee also performed a separate detailed review of the PRA to provide assurance that assessment data are traceable to reliable or controlled design documentation and information sources to facilitate future updating efforts.

The IPE review was completed in May 1996. The re.iew concluded that the study met the of Generic Letter 88-20, but icentified concerns that " post-initiator" human reliability an (HRA) yielded overly optimistic (small) HRA probabilities and that dependencies among actions were not fully considered. As stated by the licensee and illustrated by several exam m

. in the ISI submittal, the licensee screened the PRA models for initiating even was credited in the baseline PRA but is no longer feasible given the segment rupture. In cases, the probability of successful recovery was factored out of the CCDP. Removal of recovery actions ensuros that the impact of potentially optimistic HRA recovery probabiliti an initiating event caused by the failure of a segment is not a factor in the segment's categorization. For the segment failures that only failed mitigating systems, the license primarily functionallevel PRA results coupled with deterministic considerations.

An overly optimistic HRA estimate could easily place a marginally High segment two orders of magnitude before a marginally High segme l

recognizes tnat the use of overly optimistic recovery factors in the baseline PRA may minor influence on the consequence categorization of a few segments. However, the s that removal of the recovery factors that are no longer possible because a pipe seg failed, and the coupling of 'unctional level and deterministic considerations for failures, provide assuranec that any impact on the consequence categories will be m not invalidate the general results or conclusions. The licensee's analysis of possible RI-ISI specific recovery actions from the effects of segment failure are discussed in Section 3.2.6.

1 As described in the licensee's IPE submittal, the flooding analysis began with the ident the location of the equipment credited in the PRA and designation of the Appendix R fi as flood zones. A preliminary flood hazard source-location was developed, and miti features such as curbs, watertight dnors, sump pumps, floor drainage systems, and oth runoff paths were identified. A walkdown was performed to confirm the collected informatio and to verify the location (including distance above the floor level) of equipment, com and controls. The walkdown was documented with notes on scenario tab l

zone drawings. The inspection was also documented by photographing each accessible flood zone, including significant component locations, flood propagation paths, and flood mitigation / isolation features.

The results of the ISI evaluation reported in the submittal by the licensee were reviewed these parts of the IPE used extensively to support the evaluation in the submittal were iden The PRA update between the September 30,1997, submittal and the confirmatory calcu submitted in the October 8,1998 response to the staff's RADS complicated the review of th l

and its use to support the submittal. The update does, however, demonstrate the licensee's commitment to maintaining a quality PRA which corresponds to the as-operated plant. A focused review of selected parts of the IPE, supplemented by the licensee's October submittal, uncovered no shortcomings that might invalidate the results of the evaluation used support the submittal.

The staff did not review the IPE or the flooding analysis to assess the accuracy of the quantitative estimates. The staff recognizes that the quantitative results of the iPE are used as order-of-magnitude estimates for several risk and reliability parameters used to suppod the assignment of segments into three broad consequence categories. Additional estimates are used to confirm parts of the methodology whereby segments are assigned to consequence categories on the basis of the number of back-up trains available versus demand occurrence frequency categories. Inaccuracies in the models, or assumptions large enough to invalidate I

i broad categorizations developed to support RI-ISI, should have been identified in the licensee or staff reviews. Minor errors or inappropriate assumptions will only affect the consequence categorization of a few segments and will not invalidate the general results or conclusions. The staff finds the quality of the PRA sufficient to support the submittal.

Scone sf PRA Segments in the High consequence category are not further differentiated, so any segment categorized as High in the internal event evaluation need not be further evaluated for other types of initiating events.' The licensee evaluated the shutdown modertof operation and operating conditions for each of the systems included in the evaluation. The submittal contains a detailed i

l discussion of the segments involved in shutdown operational functions.

)

l' Fires and seismic initiating events are also evaluated and discussed for each system in the l

submittal. As with shutdown, only pipe segments with medium and low consequence need to be j

reviewed. In general, the licensee determined that the frequency of transients related to Class 1 l

piping and induced by external events is less than the frequency of internal events--induced l

transients. No relationship between reduced inservice inspections and increased vulnerability to l

segmer:t failure arising from the occurrence of external events was identified.

The staff finds the scope of the IPE acceptable because initiating events and operational modes outside of the scope of the IPE were systematically included in the RI-ISl evaluation and not l

neglected.

l 3.2.3 Scooe of Pioino Systems The scope of the piping systems included in the ANO-2 RI-ISI program is broader than that i

required by the Code Case or the SRP since the licensee has included the current Section XI systems as well as Code-exempt and non-Code portions of these systems, due to their potential l

safety-significance. In its response to an NRC RAI, the licensee more clearly presented the entire scope of the RI-ISI assessment, and provided the rationale for why certain systems that were analyzed in the ANO-2 PRA were not included in the RI-ISI program. For these systems, such as component and auxiliary cooling water, instrument air, and auxiliary feedwater, it was shown that loss of the systems has a negligible impact on CCDF and are therefore, of low safety importance, and do not warrant inclusion into the RI-ISI program.

l The licensee also noted that the piping scope for the RI-ISI assessment goes beyond what l

would be required by Code Case N-578. The Code Case mandates inclusion of piping within the Section XI Class 1,2, and 3 examination boundaries. However, the Code Case allows for t

piping evaluated as part of the PRA but outside the existing Section XI examination boundaries l

to be included at the owner's option. Therefore, the scope of the piping systems that were included in the RI-ISI program meets the minimum requirements of N-578, and adequately identifies the High safety-significant piping at ANO-2.

l i

r-wv4-m m

--r 3.2.4 Pioina Seaments

)

Piping systems defined by the scope of the RI-ISI program were divided into piping segments.

Pipe segments are defined as lengths of pipe whose failure lead to the same consequence and which are exposed to the same degradation mechanism. That is, some lengths of pipe whose failure would lead to the same consequences may be split into two or more segments when two or more regions are exposed to different degredation mechanism. The staff finds this appropriate, and necessary, because the methodology combines separate consequence categories with degradation mechanism categories bnd therefore the two characteristics should l

not be mixed within a segment. In some cases the licensee performed a screening analysis for welds in pipe segmenti that interfaced with a main flow path to determine if they could be characterized as Low safety significant. Safety-significance was qualitatively evaluated by assessing the direct and indirect consequences of pipe segment failure. Those pipe segments that interface with a main flow path and were found to be of low safety-significance, i.e., Risk Category 6 or 7, were cxcluded from further detailed evaluation.

The licensee's October 8,1998, submittal indicated that the following screening criteria were used:

(1)

The segment was included in an augmented program and no other degradation mechanisms identified within the segment, (2)

Segment failure (including direct and indirect effects) would not result in an initiating event and the segment is not located in the main flow path of any system performing a plant safety function that would be required to respond to a design-basis event, and (3)

The segment is normally isolated or there is a high degree of confidence that the failure would be detected and isolated because:

(a)

The room is equipped with level indicators that alarm in the control room or the room is large enough that flooding is not a hazard to equipment in the room, and (b)

The valve used to isolate the break is not affected by the break (e g., the valve is in the same room but is high enough above the floor not to be flooded and environmentally protected against spraying or jet impingement if in close proximity to the break).

A discussion is provided on the consequence degradation mechanism, isolation capabilities, and justification for piscing segments in Low for each pipe section. The staff finds that the screening process to reduce the analysis effort needed for clearly Low-safety significant segments i

acceptable because hazard and consequence information of sufficient detail to support the justification is collected, evaluated, and documented.

The SRP requires that segments prone to the same degradation mechanism be systematically identified. The licensee used an enhancement to the currently published EPRI and Code Case methodologies by assessing the degradation mechanisms in accordance with an independently written report (Structural Integrity Associates Report SIR-96-097' ). Although the criteria of the

.. EPRI guideline are very s;milar, the independent analysis presents a more refined set of degradation mechanisms with clearly defined attributes. These will be incorporated into t revision of EPRI TR-106706 and Code Case N-578 as part of the lessons learned durin application of the methods in the pilot studies. The licensee presents additionalinformation to highlight the differences between degradation mechanisms and associated attributes that considered in the Structural integrity report and the EPRI methodology.

The licensee has met the SRP requirement to confirm that a systematic process was used t identify and group pipe systems into segments with common failure consequences and susceptibility to common degradation mechanisms. Considering the supporting information presented by the licensee, the basis for defining piping segments has been adequately justif 3.2.5 Pioina Failure Potentip_I Piping failure potential was determined on the basis of the degradation mechanism evcbation as described previously in Section 3.2.4. The licensee used a methodology enhancement different from the currently published EPRI and Code Case methodologies by assessing t degradation mechanisms using Structuralintegrity Associates Report SIR-96-097". A gene desciiption of this process is included in Sections 4.1, 5.0, and Appendix B of each submittal.

In the EPRI methodology, although the consequences of piping failures are evaluated a large break, the pipe break failure potential rankings are based upon specific degradation mechanisms for which the pipe segment is postulated to be cusceptible. Only a pipe segmen that is susceptible to FAC receives a High pipe failure potential, unless that segment is susceptible to a different degradation mechanism, other than FAC, and also has the potential for water hammer. The licensee conducted a plant-specific and industry evaluation of water hammer events for applicability to ANO-2 using EPRl/NRC guidelines. The licensee conter9s that, unlike degradation mechanism evaluations, water hammer is a plant-specific phenomenon because of due to individual system configurations and operating practices. In early plant operating cycles, ANO-2 experienced a small number of these events, primarily in feedwater and main steam systems. In addition, the service water system was initially believed to be susceptible to water hammer. Plant modifications were made to eliminate condensation collection areas, and operating procedures were revised to mitigate the potential for water hammer in these systems. No events have occurred since these alterations were implemented Therefore, on the bases of plant system reviews snd an absence of recent events, the licensee has eliminated water hammer from consideration for all of the systems included in the propos alternative.

The licensee has met the SRP requirement to confirm that a systematic process was usad to identify pipe segment susceptibility to common degradation mechanisms, and to place these degradation mechanisms into the appropriate degradation categories with respect to their potential to resu't in a postulated large pipe break.

l 3.2.6 Consecuence of Failure The results of the consequence analyses of postulated pipe segment failures are presented in Sect lon 4, Table 2, and Appendix A of the submittals, except for the servica water study, which

is found in Section 5, Table 5-1 and Appendix A. As required by the SRP, t the postulated pipe failure considered both direct and indirect effects of each The direct effects always include a diversion of flow large enough to eithe lead to isolation (automatic if available, manual if feasible). Indirect effects caused by flooding, spray, and pipe whip as well as depletion of water sou tanks.

The analysis is performed assuming a large break, unless a smaller break wo severe consequences. Large breaks are generally more limiting since, for a small transient and environmental effects on the plant are less severe, and the plant op 5

more time to respond to the break. No credit is given to leak-before-break. The staf i

the use of large break, or more limiting small break, and all associated direct a to bound the impact of each break can be used to characterize the risk from l

is acceptable.

l trains) to mitigate an event. In some cases, the equi pipe rupture can vary greatly if an automatic (e.g., check valve closure and auto valves) or remote manual isolation succeeds or fails. It is necessary to represen for the operators to isolate a break and recover mitigating capability within thi framework. In general, if successful operator action to isolate of the break wou 4

more mitigating trains, the potential for isolation was credited as one backup train. C l

{

backup train corresponds, in this methodology, to a failure probability of 0.01 an i

below, the licensee provided justification for generally using a human error prob magnitude. In ocme specific scenarios,0.5 and 2 backup trains, corresponding to hu probabilities of 0.1 and 1E-4 respectively, are discussed and used in the sub successful isolation, the backup train might fail to operate, so isolation and backup fa be added. That is, failure to isolate or failure of the mitigating train given successfu would, for example, be 0.02, but this difference is negligible within the bounding valu this methodology.

actions credited in the RI-ISI evaluation. The descri 4

)

the opcrators to identify that a rupture has occurred and where the rupture is located 3

procedures and annunciator response actions directing the operators to locate and isola leak, and an estimate of the time available to recover. The license also reported an the cite in which an operator responded quickly to a reported (but, unknown to th real) leak alarm. Manual recovery was only credited when there is a control room alarm indicating leakage to which the operator will respond by investigating, each alarm directed by the corresponding procedure, and all isolation actions can be taken from t room. Rupture of one of the four HPSI lines inside containment is the exception because no leakage alarms are tripped following this event. In this case, the license described the indicating improper system operation. Response to these alarms would lead to remote isolation of the ruptured line. The staff finds that crediting isolation potential as descr submittal is acceptable because it provides for including isolation (which has a substantial impact on the consequence of pipe rupture), and the impact of not adding the recovered train's I

. failure probability to the operator error probability is negligible when compared to the magnitude analyses upon which the methodology is based.

Because of the potential for interactions between the system trains and between differ have an estimated unavailability of 2.4E-4 to provide trains with an unavailability of 0.01 each would have, at most, an unavailabi.'ity of 1E-4 Therefore, these two physical trains would then represent 1.5 backup trams. The lice performed a number of PRA calculations for the more complex failures, confirming th number of backup trains assigned to various functions was consistent with the PRA est Additionally, the licensee's October 8,1998, submittal included a systematic evaluation support system dependencies between systems and system trains to further ensure that the assigned number of backup trains was appropriate.

Each segment not screened out was evaluaten with a comprehensive failure mode an analysis (FMEA). Aside from the reactor coolant system and parts of other systems that ar located inside the containment (and designed to operate following a loss of coolant accid (LOCA)], the FMEA included a walkdown of each system. A full evaluation and discussio each segment is presented in the submittal. The description includes the direct conse indirect spatial effects, flood propagation paths, safety functions failed due to the se as well as safety functions not affected by the segment failure (used to determine the numb backup trains remaining), alarms and instrumentation arising from the rupture, and associa automatic and manualisolation possibilities. The submitted documentation indicates that the licensee identified equipment in the various areas that could be susceptible to the enviro impact of the pipe rupture and the environmental qualification of the equipment. Furthermor the flood propagation paths are identified, described, and evaluated. The staff finds that the process described by the licensee, and as illustrated by the submitted results, is acceptable allows for the development of direct and indirect effects of pipe ruptures because appr i

i information is collected, evaluated, and documented.

Consecuence Cateaarization The specific decision criteria used to determine the consequence category depend on of impact the segment failure has on the plant. In general, however, the criteria are derived fr guidelines applied to the CCDP given the segment failure. That is, given a segment failure an all the associated spatial effects, the CCDP is the probability that the resulting scenario will lead to core damage. If the failure of a segment is estimated to lead to a core damage event with a probability greater the 1E-4, the segment is categorized as High consequence. An estimated CCDP within the range of 1E-6 to 1E-4 is categorized as Medium consequence. CCDPs less than 1E-6 are categorized as Low consequences.

If the segment failure only causes an initiating event and does not cause any mitigating functions to fail, the CCDP guidelines can be directly applied. Otherwise, the methodolo provides guidance on assigning consequence category to segment breaks based on the number of available (e.g., backup) mitigating trains remaining, broad categories of initiating event frequencies, and exposure times. The guidance, in the form of a matrix, is based on the assumption that each backup train left to mitigate an event has an unavailability of 0.01. That is,

. _ _-_~..

. in order for the CCDP of the matrix elements to be assigned High, Medium, and Low to correspond to the 1E-4, between 1E-4 and 1E-6, and less than 1E-6 guidelines, each backup train must provida an availability of at ' east 0.99.

The following decision criteria are used to support the CCDP-related categorization of each type of segment failure consequence:

(1)

When the segment fal lure causes only an initiating event (e.g., no mitigating system failures caused by segment rupture), the CCDP can be estimated and directly compared i

to the guideline values.

(2)

Segment failures that only fail mitigating functions but do not cause a plant trip increase the likelihood that, following an unrelated initiating event, the sequence of events will lead i

to a core damage event. In some cases (for example, normally isolated segments), the segment failure may occur before the event but only become manifest upon demand. In.

i other cases, the failere may be detected and the repair initiated (up to the allowed outage l

time limits of the equipment), and the event may occur during the repair. The licensee uses a matrix supplied in the submittal that specifies consequence categories according to l

categories of initiat!ng events based on expected frequencies, the number of equivalent, backup trains left to mitigate the event, and exposure time. The consequence category for j

each matrix entry was developed by estimating a CCDP from the bounding values of all three contribu'ing factors, and comparing that bounding value to the CCDP guidelines.

l (3)

Segments that both cause an initiating event and fail mitigating systems are the last type of segment failure consequences. The licensee used a matrix supplied in the submittal, whereby the number of equivalent, unaffected trains availab!e for mitigation determines l

the consequence. The consequence category for these matrix entries was developed by l

estimating a CCDP assuming the bounding unavailability of 0.01 for each backup train, and comparing the result to the CCDP guidelines.

The staff finds the definition and use of the methodology and the matrices acceptable because the matrix elements are derived from bounding values, and because the licensee performed evaluations to provide reasonable assurance that each assigned backup train corresponds to at least the availability of the bounding values.

(

The licensee addresses the potential for lar0e early release for containment bypass sequences, j

and for containment failure following non-bypass core dainage sequences. The licensee does l

not provide a set of quantitative guidelines relating High, Medium, and Low safety significance to large early release likelihood. The staff has accepted an Rl-ISI proposalin which conditional large early release probabilities (CLERPs) guide!!ne values are a factor of 10 less than the CCDP values, reflecting the increased severity of release over core damage arid consistent with the difference between CDF and LERF guidelines in RG 1.174. That is, a CLERP greater than 1E-5 is a High consequence, between 1E-5 and 1E-7 is a Medium consequence, and less than 1E-7 is a Low consequence. A segment rupture is assigned the higher of the CCDP and CLERP category.

l l

. Segment failures followed by isolation valve failures that result in a LOCA outside the containment (e.g., containment bypass) are the most likely large early release contributors. In 1

the IPE, the licensee discuses the possibility of preventing core damage by depressurizing the reactor cooling system (RCS) to terminate the leak, and using shutdown cooling for long-term core cooling. Due to the generally low initiating event frequency of LOCAs outside the containment, the licensee (and most licensees) did not estimate the CCDP given such an event in the IPE. The few pressurized water reactor IPEs that did evaluate the possibility estimated CCDPs from between 0.1 and 1E-4 given that a LOCA had occurred outside the containment.

As discussed below, the licensee's methodology for categorizing these segments is generally conservative, even assuming a CCDP of 1.0 given the segment rupture and the isolation failure.

For serial motor-operated and serial check isolation valves, however, common cause failure (CCF) considerations indicate that a CCDP of around 0.1 is needed to make the CLERP consistent with the Medium category assigned by the methodology.

The licensee's methodology defines " active" isolation valves as valves that must close, and

• passive" isolation valves as valves that must remain closed. The methodology recommends that if one or less active or passive isolation valve is available to isolate a rupture, the segment failure consequence is High. If two active, or one active and one passive, valves are available, l

the recommended consequence is Medium. Two passive valves yield a Low recommended l

consequence. All other combinations yield a None consequence category. Plant-specific data j

analysis for the IPE developed estimates of 6E-3 and 3E-3 for the (active) failure to close for motor-operated valves (MOVs) and check valves (CVs) respectively. Similarly, the IPE developed estimates of 4E-7 and 4E-5 for the (passive) failure to remain closed for MOVs and CVs, respectively. NUREG/CR-5497 provides CCF estimates of 1.4E-2 for high pressure MOVs failing to close,7E-2 for a low pressure CV failing to close, and 4.5E-2 for a low pressure CV failing to remain closed (the most conservative CCFs for each scenario).

On the basis of these estimates, a segment isolated by only a single active isolation valve has a probability that the valve will not close, and that segment failure will develop into a LOCA outside containment of 6E-3 or 3E-3. Since a CCDP of almost 0.01 is needed to bring the CLERP to l

less than 1E-5. the High designation for these segments is appropriate. Similarly, the passive failure of a single isolation CV at 4E-5 is also assigned High. The passive failure of an MOV at 4E-7 is conservatively assigned High. Two active failures yield LOCA probabilities (given the pipe rupture) of 9E-5 for paired MOVs and 2E-4 for paired CVs, with other combinations being substantially lower. If the core damage probability given a LOCA outside containment is around 0.1, assigning Medium to these scenarios is, within the precision of the methodology, consistent with the accepted CLERP puidelines that Medium is between 1E-5 and 1E-7. One active and one passive failure yield LOCA probabilities of 2E-7,1E-7, and 1E-9 depending on which combination of valves is present. Assigning these scensrios to Medium is consister$t with the acceptable CLERP guidelines The maximum LOCA probability for two serial CVs failing to remain closed is 2E-6, and a CCDP of around 0.1 would also make the assignment of these scenarios as Lcw generally consistent with the accepted CLERP guidelines.

The staff finds that the methodology used by the licensee to determine that consequence category with respect to containment bypass is generally conservative with respect to the l'

of containment of around 0.1 is needed to be consistent with the acceptable CLERP guidelines.

acceptable CLERP guidelines. In three specific configurations, a CCDP given a LOCA outside I

. On the basis of the evaluations from other PWRs indicating that CCDPs following a LOCA outside containment are usually less than 0.1 and the fact that the licensee discussed a success path in its IPE indicating that a process and equipment should be svailable at ANO-2 to recover from such an event, the staff finds that the methodology's assigned consequence categories based on isolation valve failure is consistent with the intent of previously approved guidelines.

However, use of the methodology's recommended consequences may not, in genera!, meet the intent of previously accepted LERP guidelines and acceptance of this methodology for use in the ANO-2 RI-IS! analysis does not constitute acceptance of the methodology in general.

The licensee addresses the possibility of non-bypass containment failure in the November 25, 1998, submittal. The licensee stated that the IPE analysis indicated that only scenarios involving the loss of containment spray and containment cooling or containment isolation coincident with station blackout contributed to large early release. Of these sequences, only the loss of service water, which would fail both containment spray and containment cooling, had the potential to affect the categorization of the segments for ISI. All segments failures which lead to total loss of service water are categorized as High with respect to core damage. Segment failures leading to loss of one train of service water are categorized as Medium with respect to core damage, and enough margin in the CCDP exists that containment performance would not change the category. That is, the CCDP is already less than 1E-5, so aven a conditional containment failure probability of 1.0 would not cause the CLERP to exceed 1E-5. The staff finds that there is reasonable assurance that the methodology as applied by the licensee to ANO-2 is consistent with the CLERP guidelines and, therefore, acceptable.

3.2.7 Safety Sionificance Determination The safety significance of pipe segments is based on categorizing (1) the consequence of segment failure into High, Medium, or Low; and (2) categorizing the failure potential of the piping as High, Medium, or Low. Once the individual elements of risk (consequence and failure potential) are developed, they are combined in a nsk matrix that has nine elements, corresponding to various combinations of failure potential and consequence rankings.

These combinations define the basis for categorizing the pipe segments into various risk categories i through 7. Risk categories 1,2, and 3 are designated as belonging to the High safety significant group, risk categories 4 and 5 belong to the Medium safety-significant group, and risk categories 6 and 7 belong to the Low safety-significant group. The N 578 Code Case methodology requires that for Risk Category 1 the number of inspection locations be at least 50% of the total number of elements in this category; for Risk Category 2 and 3, the number of inspection locations be at least 25% of the total number of elements in each risk category; and for Risk Category 4 and 5, the number of inspection locations be at least 10% of the total number of elements in each risk category. For those segments in Risk Category 6 or 7, volumetric and surface element examinations are not required.

Quantitative uncertainty calculations are not included in the methodology. The placing of segments into broad safety-significant categories tends to reduce the sensitivity of the eventual decision on the specific values develcped from the PRA, with the exception of values near the border between the categories. The sensitivity of the values near the borders is addressed by defining a Medium safety-significance category. The Medium safety-significance category l

_...___._q l.

L

. l ensures that segments that are not clearly High or Low, will receive an intermediate level of inspection activity. The Medium safety-significance category ensures that segments whose failures which are not obviously High or Low safety-significant are treated as Medium (intermediate) severity segments, both during the final safety-significance determination and in l

the assignment of weld elements to inspect. The staff finds that the performance of quantitative uncertainty calculations would not produce information that would significantly change the results of the submittal.

The staff finds that the essignment of the safety-significance to the nine matrix elements as -

detailed in the submittal is intemally consistent and logically compelling. The staff finds that the use of the reported categories, along with other evaluation and confirmation steps detailed in this safety evaluation, provides reasonable assurance that the safety-significance of each segment is appropriately assigned.

L 3.2.8 Risk imoact of Chances The ANO-2 submittal proposes to reduce the volumetric examination of ASME Code,Section XI l

welds from 220 welds to 100 welds (the 100 inspections do not include an increase of 63 wall

)

thickness and 4 visual inspections in the service water system). In the new program, the 100 inspections have been preferentially redistributed in the High and Medium risk category i

l segments, in addition, the 220 weld examinations referenced do not include the 116 surface L

examinations that were categorically eliminated.

t I

The licensee performed a boundary calculation on the change in CDF and LERF which might be associated with replacement of the current Section XI program by the proposed RI program.

The bounding calculation only includes the potential increase in CDF and LERF that might arise from no longer inspecting those locations that would be removed from the program. Including only the potential increases, the licensee estimated an increase in CDF of 1E-8/yr. Using a similar approach, the increase in LERF due to removing inspection locations that could potentially contribute to containment bypass is on the order of 1E-9/yr.

t I

The licensee also evaluated the impact of changing the ISI program, including the risk decrease associated with the addition of new (e.g., redistributed) locations or enhanced inspections at i

some. locations in the High and Medium risk category segments. Only considering the decrease 1

in CDF that might arise from the new (e.g., redistributed to High and Medium) locations, the licensee estimated a net CDF_ decrease on the order of-1E-8/yr. Including credit for new improved inspection techniques that will be applied as part of the new inspection program as well as the new locations, the licensee estimated a CDF decrease on the order of -5E-8/yr. The licensee estimated the change in LERF by estimating that the conditional containment failure probability of 0.1 is an average, bounding value, which can be generally applied to convert CDF to LERF. Thus the licensee estimates the change in LERF as -5E-9/yr.

The staff finds the licensee's process to evaluate and bound the potential change in risk reasonable because it accounts for the change in the number and location of elements i

inspected, recognizes the difference in degradation mechanism related to failure likeHhood, and considers the effects of enhanced inspection. The staff finds that the improved inspection techniques will substantially increase the fraction of potential weld ruptures that would be l

. identified by the inspection before the flaw develops into an actual rupture. T that re-distributing the welds to be inspected with consideration of the safety plant risk receive an acceptable and often improved lev concludes that the implementation of the Rl-lSi program as described in the app be risk neutral or a risk decrease, and thus will not cause the NRC Safety G 3.3 Inteorated Decision-maklna The SRP requires that an integrated approach be used in determining the proposed R!-!Sl program by considering in concert the traditional engineering analy evaluation, and the imp!ementation and performance monitoring of piping und As noted by the licensee, the EPRI RI-ISI methodology is a process-driven app without reliance on an expert panel. However, the license applications results primarily to ensure that the process was correctly and co applied. ANO-2 made use of a multi-disciplined plant team as well as an indep review to verify the final risk results. In a response to an NRC RAI, the licensee sub further detail about reviews that were performed by the plant project team an integrated plant review team. The independent plant review did not offer an changed the pipe segment risk categorizations from application of the origin 3.3.1 Selection of Examinations The selection of pipe segments to be inspected is described in Section 7 the results of the risk category rankings and other operational considerations the number of inspections required under the existing ASME Section XI ISI progra attemative RI-ISI program. Safety-significant segments are identified in Section 8 of e submittal. With the exception of the service water system, the licensee used described in the EPRI topical report to guide the selection of examination el and Medium ranked piping segments. This requires that existing FAC programs and that where other degradation mechanisms are identified, a minimum 25% of all e within High safety-significant segments (Categories 2 and 3 - no segments were id Category 1 at ANO-2), and 10% of all elements with a Medium safety-significa (Categories 4 and 5), be examined during each interval. The EPRI report desc examination volumes (typically associated with welds) and methods of examination b the type (s) of degradation expected. The staff has reviewed these guidelines and has determined that, if implemented as described, the RI-ISIS should result in improved service-related discontinuilies beyond what is currently required by ASME Section XI.

For the service water system, the licensee elected to depart from the EPRI me because of the type of expected degradation. The piping integrity in this system l

influenced by raw water from a shallow lake, and it is expected that local degradat of microbiologically influenced corrosion (MIC), pitting, and flow-induced erosion-c dominate. Approximately 20 years of operational history validates this contention at AN The licensee has stated that for a system such as service water, the EPRI approach i l

i

. practical, because localized corrosive attack can occur within substantially large portions of the piping, and is not necessarily associated with a structural discontinuity such as a weld. For these reasons, selecting a random percentage of locations is considered arbitrary and excessive. The licensee has performed what has been termed a " finer screening," to deterrnine potential degradation sites, and to target examinations appropriately. This analysis incorporated operational parameters such as temperature, flow, water chemistry, and treatment variations, as well as results from previous monitoring and inspections, to ascertain a relative ranking of piping Table No.1 S

W L afetyjAs, MASME Coder 4.j Curant Q (Proposed RI t einspection Nos.

. Significance ~ X@ Class A kSectionXI6 M idISIe

• .) Changed 1

, y a @4

Inspections ?

ilaspectionsm3 e

1 33 28

-5 2

23 6

-17 High 3

0 20

+20 l

Non-class 0

2

+2 1

101 40

-61 l

2 19 27

+8 l

Medium 3

L 37

+37 Non-class 0

0 0

1 59 0

-59 2

101 0

-101 Low 3

0 7

+7 Non-class 0

0 0

l l

(1) New inspections are in addition to existing flow-accelerated corrosion (FAC) program.

l l

t j

subsystems within the service water system, in terms of expected susceptibility to localized corrosion. The risk categorization performed previously (High, Medium, or Low) was not initially considered during this susceptibility determination; therefore, a list of examination locations within various segments from all risk categories resulted. However, piping locations rated as

" worst case' were subsequently screened to determine if any High safety-significant (based on

. _ _. - =. _ _ _. _

t i

19-failure consequence) elements with similar degradation susceptibility might be substituted for l

any Medium and Low safety-significant segments that resulted from the susceptibility review.

This approach resulted in the selection of 67 " typical" and " worst case" locations for examination. These examinations will be performed during each inspection period, because of the aggressive nature expected for localized corrosion phenomena. In addition,26 separate i

locations will be inspected every 2 years in accordance with the Service Water Integrity Program, a licensee-controlled augmented program currently in place at ANO-2. The licensee j

has also described conditions under which the 67 locations would be expanded, based primarily on the guidelines in NRC Generic Letter 90-05.

It is noted that the licensee's distribution of examination locations does not entail all piping segments in High and Medium safety-significant categories. Forinstance, of 8 High safety-significant segments in the service water system,20 of the 67 examination locations will be performed on only 4 of the piping segments. However, on the basis of the expectation that degradation will be initially manifested in the " worst case" locations, and because a sample expansion method has been described, it is concluded that generic degradation due to localized corrosion of service water piping will be detected and the scope of affected piping will be adequately defined, to provide reasonable assurance of the continued structuralintegrity of this system. Further, it is expected, that, if new potential degradation information is found as a result of ANO-2 specific data or other generic industry issues, the licensee will factor these findings

'into the current program for the service water system, as described in Section 3.4, which follows.

However, any change that would reduce the number of examinations will require review and approval by NRC staff.

For the service water system, it is concluded that the degradation susceptibility review process, augmented with the selection of higher risk locations for those locations with equivalent susceptibility rankings, has resulted in a reasonable method for establishing a program to assess service-induced degradation caused by localized corrosion. Further, for the remaining systems, it is concluded that the overall risk-ranking process has resulted in the systematic identification of safety significant pipe segments, and the approach found in the EPRI methodology provides adequate justification for the locations to be examined.

3.4 Imolementation anGonitenng Implementation and performance monitoring strategies require careful consideration by the licensee, and are addressed in Element 3 of the SRP. The objective of Element 3 is to assess performance of the affected piping systems under the proposed RI-ISI program by implementing monitoring strategies that confirm the assumptions and analysis used in development of the RI-ISI program. To satisfy 10 CFR 50.55a(a)(3)(i), implementation of the RI-ISI program, including inspection scope, examination methods, and methods of evaluation for examination results, must provide an adequate level of quality and safety. In a response to an NRC RAI, the licensee has indicated that " implementation of the ANO-2 RI-ISI program will be consistent with existing ASME Section XI performance monitoring requirements," including pressure and leak testing of all Class 1,2, and 3 piping components, comparison of inspections results to PSI and prior ISI, and adherence to IWX-3500 for flaws that exceed acceptance criteria.

w w

e l

l l As stated by the licensee, "An inspection for cause process shall be implemented utilizing l

examination methou, and volumes defined specifically for the degradation mechanism l

postulated to be active at the inspection location." The examination methods and volumes are to be based upon the requirements defined in Section 7 of the EPRI methodology. In order to support a finding that the proposed alternative is acceptable, the licensee has provided in its responses to NRC RAls the following items:

Effective January 1,1998, Entergy at ANO uses only UT personnel that meet the qualification i

requirements of the 1992 Edition of ASME Section XI, Appendix Vil, and only UT personnel who have successfully completed the PDI qualification will be used for ultrasonic examinations.

Inspection intervals are based on the current Section XI 10-year inspection interval, supplemented with augmented inspections for specific degradation mechanisms based on NRC-mandated inspection schedules or the plant's own program.

Inclusion of vessel nozzle dissimilar metal welds is in the scope of the Rl-ISI program.

Monitoring of industry trends to assure that identification of new degradation mechanisms or component susceptibility to existing mechanisms will result in changes to the RI-ISI program to incorporate these changes, as well as any changes mandated by the NRC and industry owners groups.

l In consideration of the implementation and monitoring program proposed by the licensee, the reliability and monitoring of the examinations to be performed under the RI-ISI program is acceptable.

4.0 CONCLUSION

S in accordance with 10 CFR 50.55a(a)(3)(i), proposed alternatives to regulatory requirements may be used when authorized by the NRC when the applicant demonstrates that the alternative L

provides an acceptable level of quality and safety. In this case, the licensee's proposed alternative for piping systems is to use Code Case N-578, augmented by EPRI TR-106706, and the methodology enhancements discussed herein. These enhancements include commitments by the licensee to do the following:

Use ultrasonic testing (UT) personnel who meet the qualification requirements of the 1992 Edition of ASME Section XI, Appendix Vil by successfully completing the industry's Performance Demonstration initiative qualification.

i Continue to implement augmented examinations in the areas of (1) flow accelerated corrosion (FAC) and (2) classification of "no-break zone" examinations in accordance with MEB 3-1.

Use expanded examination volumes as described by the methodology in EPRI Topical Report-106706.

l

. The alternative is documented in the licensee's RI-ISI program submittals, supplemented by the licensee's responses to the NRC's RAls. On the basis of the review of these documents, it is concluded that the licensee's risk-informed approach should result in a risk-neutral to risk-reduction status when compared to the current ASME Section XI program, and should significantly reduce the number of examinations performed.

In addition, the licensee has met the applicable criteria in SRP Chapter 3.9.8. Therefore, on the basis of risk considerations and the criteria of the SRP, it is concluded that the licensee's proposed alternative to use Code Case N 578, with the specific augmentations described in this report, will provide an acceptable level of quality and safety. Therefore, the proposed alternative is authorized for use at ANO-2. It should be noted that, although the licensee's alternative includes the use of ASME Code Case N-578, this evaluation is not endorsing the Code Case as currently written. This evaluation applies only to piping systems at ANO-2; all Code-required inspections of other safety-related components shall continue to be performed in accordance with ASME Section XI and as required by the licensee's Technica! Specifications. The use of this alternative is authorized for the remaining license term of the plant. However, any deviations that would decrease the scope of examination, or change the overall plant risk will require staff review and approval.

5.0 REFERENCES

1.

Letter, dated September 30,1997, D. C. Mims (Entergy Operations, Inc.) to Document Control Desk (NRC), containing results of pilot plant study for risk-informed ISI program at ANO-2.

2.

Letter, dated March 31,1998, D. C. Mims (Entergy Operations, Inc.) to Document Control Desk (NRC), requesting approval of a risk-informed alternative for examination of piping systems at ANO-2.

3.

Letter, dated October 8,1998, J. D. Vandergrift (Entergy Operations, Inc.) to Document Control Desk (NRC), containing additional information in support of risk-informed ISI pilot application at ANO-2.

4.

Letter, dated November 25,1998, J. D. Vandergrift (Entergy Operations, Inc.) to Document Control Desk (NRC), containing additional information in support of risk-informed ISI pilot application at ANO-2.

5.

Letter, dated December 8,1998, J. D. Vandergrift (Entergy Operations, Inc.) to Document Control Desk (NRC), containing additional information in support of risk-i'. formed ISI pilot application at ANO-2.

6.

NRC Regulatory Guide 1.178, An Approach for Plant-Specific, Risk-Informed Decision Making: Inservice inspection of Piping, Issued for Trial Use, U. S. Nuclear Regulatory Commission, September 1998.

7.

Standard Review Plan (SRP) Chapter 3.9.8, Standard Review Plan for Triel Use for the Review of Risk-Informed inservice Inspection of Piping, U. S. Nuclear Regulatory Commission, September 1998.

.9

~

1 j

-4 8.

EPRI TR-106706, Risk-Informed Inspection Evaluation Procedure, Interim Report, Electric Power Research Institute, June 1996.

9.

ASME Code Case N-578, Risk-Informed Requirements for Class 1, 2, and 3 Piping, j

Method B,Section XI, Division 1, American Society of Mechanical Engineers.

' 10.

SIR-96-097, Review of Degradation Mechanisms in EPRI Risk Informed Inservice Evaluation Procedure, Structural lntegrity Associates, Inc., November 1996.

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'I J

i Principal Contributors: Stephen Dinsmore Tom McLellan Date: December 29, 1998 d

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