1CAN059902, Responds to NRC 990406 RAI Re risk-informed Inservice Insp Pilot Application,Submitted 980603.Approval of Alternative Is Requested Prior to End of July 1999,to Allow Sufficient Time for Util to Revise ANO-1 ISI Program

Responds to NRC 990406 RAI Re risk-informed Inservice Insp Pilot Application,Submitted 980603.Approval of Alternative Is Requested Prior to End of July 1999,to Allow Sufficient Time for Util to Revise ANO-1 ISI Program

ML20206U165
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/17/1999
From:	Vandergrift J ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN059902, 1CAN59902, NUDOCS 9905250077
Download: ML20206U165 (57)

Text

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O -- Ent;rgy Operattions. Inc.

${j h[" 1448 SA 333 i RussdMte, AR 72801 Tal 501858-5000 1

May 17,1999 1CAN059902 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station OPI-17 Washington, DC 20555 l

Subject:

Arkansas Nuclear One- Unit 1 Docket No. 50-313 .

License No. DPR-51 Additional Information in Support of Risk-Informed Inservice Inspection Pilot Application Gentlemen:

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Entergy Operations submitted the results of the risk-informed inservice inspection (ISI) pilot plant application study for Arkansas Nuclear One, Unit 1 (ANO-1), to the NRC by letter dated June 3,1998 (ICAN069804). The pilot plant application is to be used as an i alternative, per 10CFR.50.55(a)(3)(i), to certain ASME Code requirements for the remainder of ANO-l's third inspection interval and all subsequent inspection intervals. A request for l additional information from the NRC staff was received on April 6,1999 (ICNA049903).

The requested information is attached.

As discussed in the June 3,1998, letter, a risk-informed inservice inspection is planned for the next ANO-1 refueling outage (IRIS), currently scheduled to commence on September 10,1999. Approval of this alternative is requested prior to the end of July 1999, i to allow sufficient time for Entergy Operations to revise the ANO-1 ISI program to i incorporate the risk-informed approach prior to its implementation.

Should you have any questions regarding this submittal, please contact me.

G 1: 18 9905250077 990517

PDR ADOCK 05000313 G PDR l

U. S. NRC May 17,1999 1CAN059902 Page 2 Very truly yours, r r im D. Vander or, Nuclear ety JDVfjd attachment cc: Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission RegionIV  !

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 Mr. Nick Hilton NRR Project Manager Region IV/ANO-1 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-D-18 One White Flint North 11555 Rockville Pike Rockville, MD 20852 l

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Attachment to 1CAN059902 Page1of55 Entergy Operations Response to NRC Questions on Risk-Informed Inservice Inspection Pilot Plant Application NRC Question No.1 Regulatory Guide (RG) 1.174, as well as Standard Review Plan (SRP) 3.9.8, identines nye preciples of risk-informed (RI) regulations. They are:

a) Meet current regulations unleu explicitly related to a requested exemption or rule change, b) Consistent with defense la depth philosophy, c) Malatala sufncient safety margins, d) Proposed increase la core damage frequency (CDF) or risk are small and are consistent with the Commission's Safety Goal Policy Statement (See also ,

Question 9), and e) Use performance measurement strategies to monitor the change Please explain how the proposed change to the Arkansas Nuclear One (ANO-1) inservice inspection (ISI) program meets the above listed five principles. Principle )

(d) on the potential risk increase can be answered la Question 9. l l

Entergy Operations Response to la

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i 10CFR50.55a and Appendix A to 10CFR50 are the primary regulations governing inservice inspection of piping. The intent of these regulations, as it pertains to the Code Case N-560 scope of piping, is to ensure a robust reactor coolant pressure boundary.

Through 10CFR50.55a, Section XI to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code is the implementing vehicle for these inspections. Code Case N-560 is an ASME approved alternative to current Section XI requirements.

Other Section XI inspection activities such as the examination of Class I socket welded connections and Category B-F dissimilar metal welds, pressure and leak testing requirements, and Class 2 and 3 piping examinations, are not affected by implementation ofN-560.

In addition, certain augmented inspection programs such as those implemented in response to Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," and ' Generic Letter 89-08 , " Erosion / Corrosion Induced Pipe Wall Thinning," are also not impacted by implementation of N-560. A full accounting of the impact of the N-560 application on current ANO-1 augmented inspection programs is provided in response to Question 18.

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Attachment to ICAN059902 Page 2 of 55 '

Other plant programs, which are not inspection related, but can have a dominant impact on assuring piping reliability, include the primary water chemistry control program, reactor coolant leakage, containment temperature, pressure, and radiation monitoring, all of which are unaffected by implementation of Code Case N-560.

Entergy Operations Response to Ib The piping systems at ANO-1 contribute to defense-in-depth in two important ways: The piping of the reactor coolant system (RCS) and systems that directly interface with the RCS provide one of the sets of barriers in the barrier defense-in-depth arrangement. This barrier protects the release pathway from the reactor core to containment release pathways and part ofit is responsible for protecting against potential containment bypass pathways. The second way piping contributes to defense in depth is its role in the protection of the core through providing critical safety functions that are dependent upon piping system integrity.

The role that inspection programs at ANO-1 play in determining the risk significance of piping systems is rather limited and well defined. The piping inspections play a role in identifying defects and degradation in piping system elements. When defects and degradation damage are found and repaired, pipe failures are precluded and the probability of pipe rupture reduced. In addition, pipe inspections and leak tests and detection processes have the potential of correcting pipe problems and reducing the safety function unavailability's due to pipe failures. Hence, the impact of changes in the inspection program are limited to potential changes in failure frequency and mpture frequency, but do not change the consequences of an assumed pipe failure.

As discussed above, the intent of the inspections mandated by ASME Section XI for category B-J piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or rupture in the reactor coolant pressure boundary. Currently, the process for picking inspection locations is based upon structural discontinuity and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev.1, " Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds in Piping," and Electric Power Research Institute (EPRI) topical report TR-112657,

" Revised Risk-Informed Inspection Evaluation Procedure," this method has been ineffective in identifying leaks or failures. In response to these findings, ASME issued Code Case N-560, which has a much more focused selection process founded on actual service experience with nuclear plant piping failure data.

The N-560 selection process has two key ingredients. Those are: 1) a determination of each location's susceptibility to degradation, and 2) an assessment of the consequence of  ;

the location's failure. These two ingredients not only assure defense-in-depth is maintained, but is actually increased over the current process Initially, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures in the reactor coolant pressure boundary is

e Attachment to 1CAN059902 Page 3 of 55 increased. Secondly, the consequence assessment effort has a single failure criterion so that, no matter how unlikely a failure scenario is, it is ranked high if, as a resuh of the failure, there is no mitigative equipment available to respond to the event. In addition, the consequence assessment takes into account equipment reliability so that poor performing equipment is not credited as much as more reliable equipment.

Entergy Operations Response to Ic The safety function ofinterest in the ANO-1 RI-ISI evaluation is that of system pressure boundary integrity. Listed below are those attributes necessary for fulfilling this requirement, as well as the impact of the ANO-1 RI-ISI program on meeting the i objective: I Quality Design - No Change Quality Fabrication - No Change Quality Construction - No Change Quality Testing - No Change Quality Inspection - Fewer inspections conducted at more appropriate locations using  ;

better techniques and as necessary expanded volumes.

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As can be seen from the above summary, those attributes that are critical in defining and maintaining sufficient safety margins are unchanged, except for a subset of the pressure boundary volumetric examinations. In this case, the reduced number of volumetric Section XI exams is based upon an explicit consideration of potential degradation and the accompanying consequence of system failure. As such, the new Section XI inspection locations are more appropriate and will have enhanced inspections conducted.

Entergy Operations Response to 1d See response to Question 9.

Entergy Operations Response to le Implementation of the ANO-1 RI-ISI program will be consistent with existing ASME Section XI performance monitoring requirements. These are as follows:

e pressure and leak testing of all Class 1 piping components e inspection results shall be compared to preservice inspections and prior ISI e for flaws exceeding mem criteria (IWB-3500) l l

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Attachment to ICAN059902 Page 4 of 55 increase the number ofinspections to include those items scheduled for this and the next scheduled period additional flaws - all items of similar design, size and function 1- flaw - removed, repaired, replaced or analytical evaluation flaws accepted by analytial evaluation shall be examined for the next three inspection periods In addition to the above ASME Section XI monitoring and feedback mechanisms, l ANO-1 has in place the following processes to detect reactor coolant leakage as I described in section 4.2.3.8 of the ANO-1 Safety Analysis Report:

sump level inventory balance radiation monitoring

Attachment to ICAN059902 Page 5 of 55 NRC Question No. 2 The licensee should identify or provide a cross-reference between the risk segment identifiers in Tables 3-1,3-2 and 3-3 of Report No. SIR-98-055, " Risk Evaluation and Element Selection in Support of ASME Code Case N-560, ANO-1," and the line number / segment identifiers la Table 4-1 for the consequence category rankings in l Calculation No. NSD-024, "ANO-1 N560 Consequence Evaluation," to facilitate review of the flaal exam locations that were selected.

Entergy Operations Response J

The following table provides the requested cross-reference between Table 4-1 of Calculation No. NSD-024 and Tables 3-1,3-2 and 3-3 of Report No. SIR-98-055. The l

conesponding risk segment (s) are provided for each line number / segment in the order j that they appear in the respective risk segment tables of SIR-98-055. j i

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g 0 Attachment to ICAN059902 Page 6 of 55 NSD-024 / SIR-98-055 CROSS REFERENCE TABLE

Line Numbers /h=fets froan Table 4-1 of Corresponding Risk Segments from Tables -

' Calculation No. NSD-024 -

' 3-1,3-2 and 3-3 of Report No. SIR-98-055 :

CCA-1-36 RCS-001-1, RCS-002 CCA-1-28 RCS-003, RCS-004-1, RCS-005, RCS-006-1, RCS-007, RCS 008-1, RCS-009, RCS-010-1 CCA-2-10 RCS-001-2, RCS-011-1, RCS-011-2, CCA-3-2.5 upstream of CV-1213 and CV-1215 MU&P-005-1, MU&P-005-3, MUAP-005-5, MUAP-005-11, MU&P-005-2 (Includes 1% inch elements24-065 and 24-067- See notefor consequence segment "CCA-13-1.5 upstream of mlves RBD-BA,B,C,D"), MU&P-005-4, MU&P-005-10 CCA-3-2.5 downstream of CV-1213 and CV-1215 MU&P-005-6, MU&P-005-8, MU&P-005-12, MU&P-005-14, MUAP-005-7, MU&P 005-9, MU&P-005-13, MU&P-005-15 CCA-4-2.5 & 4 (unisolable) RCS-012-2, RCS-012-4, RCS-012-5, RCS-012-1 CCA-4-2.5 (isolable pipe between CV-1008 and 1009) RCS-012-3 CCA-5-2.5 downstream of check valve MU-34A,B,C,D RCS-004-2, RCS-006-2, RCS-008-2, RCS-010-2, MU&P-001-5, MUAP-001-7, MU&P-002-2, MU&P-003-5, MU&P-003-7, MU&P-004-5, MU&P-004-7 CCA-5-2.5 upstream of check valve MU-34A,B,C, or D MU&P-001-4, MU&P-001-6, MU&P-001-8, MUAP-002-3, MU&P-003-4, MU&P-003-6, MU&P-003-8, MUAP 004-4, MUAP-004-6, MUAP 004-8, MU&P-001-2, MU&P-001-3, MUAP-003-2, MU&P-003-3, MUAP-004-2, MU&P-004-3, MU&P-001-1, MU&P-002-1, MU&P-003-1, MU&P-004-1 CCA-6-14 downstream of check valve DH-14 A,B DHR 0013, DHR-002-3, DHR-001-2, DHR-002-2 CCA-6-14,12,&8 upstream of check valve DH-14 A or B DHR-003-4, DHR-001-1, DHR-002-1 CCA-8-12 upstream of motor operated valve (MOV) DHR-003-1, DHR-003-3, DHR-003-2 CV-1050 CCA-8-12 downstream of MOV CV-1050 DHR-003-5 CCA-9-1.5 downstream of check valve DH-12 RCS-012-6 CCA-9-l.5 upstream of check valve DH-12 No corresponding risk segments. This section of line is socket welded andis outside the scope ofthe N-560 assessment.

CCA-13-1.5 upstream of valves RBD-8A,B,C,D RCS 013-1, RCS 013-2, RCS-013-4, RCS-013-5, RCS-014-1, RCS-014-2 (Note: Two 1 % inch piping elements upstream ofRBD-8A were included in MU&P-005-2 as noted abovefor consequence segment "CCA-3-2.5 upstream ofCV-1213 and CV-1215")

CCA-13-1.5 & 2 downstnwn of valves RBD-8A,B,C,D RCS-013-3, RCS-014-3

Attachment to 1 ICAN059902 Page 7 of 55 NRC Question No.3 Section 5.0. "Results," of NSD-024, "ANO-1 N560 Consequence Evaluation,"

indicates that reactor coolant piping falls into "High" consequence category because a pipe segment failure causes a loss-of-coolant accident (LOCA) with a conditional core damage probability (CCDP) la the high range. Section 2.2, " Impact Group Assessment," also states that,"[t]he evaluation is performed such that pipe segments can be qualitatively biased using the bias in Table 2-4 (e.g., High) and/or by their quantitative CCDP value estimated la this evaluation. This allows additional ranking to be performed withis a category (e.g., High)."

Since both small and large LOCA's are ranked as having High consequences, does this statement imply that further ranking withis the High consequence segments or during element selection was performed based on the quantitative determination of CCDP7 If so, please describe how any additional ranking process was applied in the analysis.

- Entergy Operations Pasponse The ANO-1 N-560 submittal is one of a number of pilot plant specific applications of the EPRI RI-ISI methodology. To this end, Entergy and EPRI conducted additional analyses and evaluations in support of the development and application of the EPRI RI-ISI methodology beyond what is anticipated to be nace y to support future N-560 submittals. The additional ranking discussed in NSD-024 was not used to further rank the "high" segments or in the element selection process, but rather was used by the EPRI team for internal benchmarking purposes.

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Attachment to 1CAN059902 Page 8 of 55

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NRC Question No. 4 Code Case N-560, Section 2.6, requires that examination zones shall be selected starting with the structural elements la the High risk gmup and working toward the

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Law risk group, until a total number of structural elements equal to 10 percent of the Category B-J weeds, excluding socket welds, has been selected. Report No.

SIR-98-055 indicates that there are 394 total elements la the Class 1 piping at .

ANO-1, and thus,10% would be 40 welds. The examination locations proposed for )

the ANO-1 RI-ISI program consists of 40 welds divided between 28 Highs and 12 i Mediums. Since there were 84 High risk elements identified but only 28 selected, -

does this selection constitute a departum fmm the N-560 Code Case guidance? If j so, the licensee should pmvide the rationale as to why examination locations were I chosen based on belonging to the Medium risk gmup instead of the High risk group.

Entergy Operations Response As stated in the request for additional information (RAI), the number of elements to be examined as part of the RI-ISI program is defined in ASME Code Case N560 as 10 percent of the piping weld population. Guidance on choosing those locations to be '

selected for examination is provided in the EPRI repon, " Revised Risk-Informed Inspection Evaluation Procedure," TR-112657, Section 3.6.5. This guidance was adhered to during performance of the ANO-1 application. A portion of Section 3.6.5 is provided as follows:

ASME Code Case N560 repires that a 10 percent .manpling size be weighted to the higher risk locations. Provided a segment is depned as a continuous run of piping susceptible to the same degmdation mechanism (s) and same consequence offailure, it is not always necessary to inspect every location within a segment in order to capture the risk associated with that segment. In addition, many times there will be runs ofpiping defned as discrete segments but essentially identical infailure and consepence potential. Thefollowingprovides an example ofhow j a portion of a segment's locations was chosen and the remaining inspections 1 allocated to other segments, thereby increasing the portion of risk addressed by N560.

4 For exangpie, for a population of 500 welds a sample size of 50 locations is l repired in Feedwater systems ofBWRs, the horizontalsections connected to the i I

reactor vesul may be subjected to thermal stratipcation. These piping sections will generally be classiped as risk category 2, assuming a combination of high consepence (large LOCA) and mediumfailure potential (thermalfatigue, TF).

Assumig 4 locationsper horizontal sections and 4feedwater lines, a total of16 l inspection locations exist. Instead of examinig all 16 locations, it would be consideredmore prudent to examine, say six of these locations, ami use the other ten inspections (16 minus 6),for other segments anfor degradation mechanisms.

Attachment to 1CAN059902 Page 9 of 55

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With respect to the ANO-1 submittal, the decay heat removal (DHR) system is utilized as an example. The following information is taken from Table 3-3 of SIR-98-055:

i Segment ID Description Damage # of # of N560 Mechanism Elements Inspections DHR-001-3 Low Pressure Injection (A) TT/rASCS - 1 5 2 TASCS-4 DHR-002 3 Low Pressure Injection (B) TT/fASCS - 1 6 2 TASCS-5 DHR-003-1 DHR Suction Line PWSCC 1 1 DHR-003-3 DHR Suction Line TASCS 4 1 Totals 16 6 TT = thermal transients TASCS = thermal stratification PWSCC = primary water stress corrosion cracking As indicated, the DHR system consists of four risk category 2 segments that were identified as potentially subject to thermal fatigue (TT and/or TASCS) or susceptible to PWSCC.

Two locations in segment DHR-001-3 were chosen for inspection, of which one of these locations includes the location potentially subject to both TT and TASCS. Similarly, for segment DHR-002-3, two locations were chosen for inspection including the location potentially subject to both TT and TASCS. For the single element segment, DHR-003-1, one of one location is chosen for inspection. For the remaining segment, DHR-003-3, I one of four locations is chosen for inspection. f l

By inspecting six of these sixteen locations, each of the potential degradation )

mechanisms identified for the DHR system is addressed. In addition, this allows the j remaining ten inspections to be allocated to other areas thereby capturing additional risk j and/or potential degradation mechanisms. Defining segments as runs of piping susceptible to the same degradation mechanism and consequence of failure is important to this inspection strategy.

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Attachment to ICAN059902 Page 10 of 55 NRC Question No. 5 In order to conclude that the results and conclusions reRect the as-built and as-operated plant, please provide a discussion on the fouowing.

a. Code Case N 560 (a)(4) requins that the consequence ranking be performed by a multidisciplinary teams of experts. Experience with the Maintenance Rule and other risk-leformed applications has shown that the expert panel should be composed primarily of plant personnel to ensure that sufHcient plant specific experience is available to provide confidence that the results renect the as built, as operated plant. ANO-l's submittal appears to have been primarily developed by two contracton and them was no mention of review and approval by a team of experienced plant persosseL The licensee should describe the makeup of the ISI selection team for the ANO-1 analysis and a description of the process and rationale that was used to classify pipe failure potential and final element exam location selection. In addidon, please describe the ANO-1 plant team which ny ewed 8 the consequence analysis work and approved the results and conclusions.

b. Your submittal refemaced a "Rev.1" of its probabilistic risk assessment (PRA), but the CDF referenced la Section 3.1 is from a Table la the licensee's April 1993 ladividual plant examination (IPE) submittal. Also, the IPE submittal cover letter indicated the large, early release fraction (LERF) was 4

expected for 2.8 percent of the CDF, or 1.3x10 /yr. Please supply the date of the last PRA model up-date and briefly describe how the up-date was reviewed and approved. If the last review of plant data and plant changes determined that a PRA up-date was not necessary, the date of this review may be considered the last PRA model up-date. The corresponding baseline CDF and LERF should also be provided to further define and characterize the PRA model used.

c. The staff evaluation report on the IPE identified potential weakness in the human error evaluation process. Specifically, the report stated that plant specific performance shaping factors were not applied and there were no walkdowns performed to support the response times obtained from operator interviews. Please clarify how these issues could affect the results and conclusions reported in the submittal, and describe any evaluations performed during the RI-ISI analysis to ensure that any affects are minimized such that the results and conclusions are valid.

Entergy Operations Response to 5a Code Case N-560 is currently undergoing revision by ASME Section XI to incorporate lessons learned from the pilot applications (i.e., Vermont Yankee and ANO-1). This revision includes changes concerning the ISI Selection Team and the process for

o Attachment to ICAN059902 Page 11 of 55 conducting the RI-ISI evaluation. These revisions were made to reflect the process actually employed in the pilot applications.

The present N-560 requirement to assemble an ISI selection team to perform the ranking process, as delineated in (a)(4) of the Code Case and I-1.0 of Appendix I, is being modified in N-560-2 as indicated below. The proposed process is consistent with the l approach employed in Code Case N-578. This proposed revision was approved by the Working Group on Implementation of Risk-Based Examination and the Subgroup on Water-Cooled Systems at the recent February 1999 Code meetings, and will be considered by the ASME Section XI Subcommittee at the May Code meetings.

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Attachment to 1CAN059902 Page 12 of 55 N-560 with Proposed N-568-2 Changes Indicated N-568-2 with Proposed Changes Incorporated (4) ^_^ - '; his process shall be (4) His process shall be performed in performedjb = '"' "'* ^'-- T-- - " la accordance with accordance with Appendix 1. De expertise required to Appendix L De 6 expertise conduct this process, shall include knowledge of the function i 6 _ :- - : d S " ," - - '- ' 4e-4be-salesteen and operation of the system, Probabilistic Safety Annaamment ]

renuired to condur.1 tins process, including-aball inchade (PSA), failure pasannal. metallurgy, stress analysis and knowledge of the functaan and operation of the system, knowledge of existing preservice and inservice examination Probabilistic Safety Assessement (PSA), failure potentant results.

metallmsy, stress analysis and imowledge of exisung p,me,vice and inservice e., i.maa. ,e. uhs.

1 m I-1,9 INTRODUCTION Dis Appendax provides the risk-informed process to be '

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used as an akernative to current selection and inspection

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, . __ T n_gr 71- . E '- requirements for piping that will be schariniarl for inservice

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- _ .u ' - m _ _ _, .u _ - .2 inspareaa at a nuclear power plant. nis alternative

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selection process is based on the relative risk-siaain- of Z( ; J - . _ _ _ - - _ _ , _ '., _ .

locahons within an individual system. Fig.1-1 illustates the

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evalustaan process that is sununarized below.

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and idannine=tia= of system boundanes and functions.

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I-1.2 Segeneet Risk Ama*=ma=='

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. . _ _ z _ _ J_2 _ . 7 T E"". _ -~ Dis step divides each selected system into piping

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segments that are determmed to have similar consequence of

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'd 6 failure and pasa=*i=1 for failure (common degradation

=_+ --- etc.). Dese segments are then placed into risk

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~ -6 ' categories which are defined based on combinations of 2.1 __- ,,_

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_2, i - ]' - {- f_. nanaaquence and failure potential Once categorized, the risk-signaficant segments are identified.

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Dis step identafies potential locations (elements) to be lampeaad within the risk-significant segments based on the specific degradation mechanism identified in the segment.

I-1.4 Inspection Imentions and Ennenlaation Methods nis step selects the elements that will be included in the ISI program. De inspadian voluanc and method used for the alamassa are deternuned based on the degradation  !

machansam amanciatant with the element.

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ne results of this pKad s (scope, input, output) must be fully darnmensatt and reviewed. His evaluation shall meet the Owner's Quality Assurance program and shall include a multi <liscipline review in% .ilug plant specific and industry experience as well as the results of current ,

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Attachment to 1CAN059902 Page 13 of 55 The EPRI RI-ISI methodology is a process driven approach. The ANO-1 staff has reviewed the application results to ensure that the process has been correctly and comprehensively applied.' A complete and thorough application of the EPRI process results in an accurate risk categorization of piping segments. Consequently, the EPRI methodology is not reliant upon an expert panel or "ISI Selection Team" to identify risk 4 significant segments that the process may have failed to identify. I The review process employed for the ANO-1 N-560 RI-ISI application was very similar to that employed for the ANO-2 N-578 RI-ISI application as documented in the response Question 30 in the request for additional information for ANO-2 (October 8,1998, 2CAN109801). The review process utilized in the ANO-1 pilot application consisted of 1 the following:

• Duke Engineering & Services (DE&S) and Structural Integrity Associates (SIA) functioned as the principal investigators for the ANO-1 pilot project in the application of the EPRI RI-ISI methodology. The consequence (DE&S) and degradation mechanism (SIA) evaluations were perfonned and documented in the form of calculations. As such, after preparation of the calculations, an independent review was performed and all comments resolved before the calculations received final approval. The risk evaluation and element selection report was prepared by Entergy Operations' personnel and documented as an SIA report.

• To support the project, Entergy Operations assembled a dedicated team of ANO plant staff with diverse experience in probabilistic risk assessment, operations, and inservice inspection. The ANO staft's duties included the following: collection of required design inputs, responding to inquiries from the project principal investigators, resolution of issues arising during the evaluations, review of plant service history relative to piping pressure boundary degradation occurrences, and a comprehensive review of the calculations with primary focus on the technical accuracy and completeness of the consequence and degradation mechanism evaluations. This team, which interfaced with other plant staff personnel on an as needed basis, was an active participant in the application of the EPRI RI-ISI process, and ensured that plant specific knowledge and insights were factored into the risk evaluations.

The ANO staffis ultimately responsible for the technical content of the ANO-1 pilot '

application submittal. ANO Procedure A4.502, " Accuracy of Communications,"

provides a process to document that complete and accurate information is submitted to the NRC. The process requires that the originator compile the documentation necessary to substantiate that the information is true, accurate, and complete. A verification .is then performed to confirm that the documentation supports the statements made. The ANO staffperformed this function.

Attachment to ICAN059902 Page 14 of 55 As stated above, application of the EPRI prccess driven approach will yield appropriate results. Application of this approach obviates the need for an expert panel or "ISI Selection Team" to ensure the appropriate risk categorization ofpiping segments.

This point is illustrated by the following two key elements of the EPRI process driven approach.

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e Application of the process ensures the identification and inspection of high j consequence / low failure potential piping segments (i.e., Medium Risk - Category 4)

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e Application of the process ensures the identification and' inspection of low consequence / high failure potential piping segments (i.e., Medium Risk - Category

5) 1 Both of the above are accomplished without the reliance on an expert panel. To further l

illustrate this point, this approach is contrasted with the following results from the Surry pilot application. j I

e 50% of the high safety significant segments were identified by the expert panel I e These segments had originally been determined to be oflow safety significance via j the risk importance measure calculations e These segments were reassigned by the expert panel to be of high safety significance due solely to high consequences in the event of failure The two RI-ISI approaches (i.e., N-577 and N-560/N-578) simply employ different means to reach the same end. Whereas the N-577 approach is reliant upon an expert panel for the identification of such segments, the N-560/N-578 approach is not. The process itself ensures the identification of all risk significant segments.

Entergy Operations Response to Sb The ANO probabilistic safety assessment (PSA) plant models are being maintained as "living models." The IPE based PSA model has been superceded by a later model that is based upon the ANO-1 plant (i.e., plant procedures and hardware) as of September 14,1993. The latest model results show that the RI-ISI consequence evaluation results are conservative. This is discussed in more detail in the response to Question 6a.

This latest model was updated by performing a review and update of each of the following elements of the model: 1) System analysis work packages; 2) The success criteria and top logic for the each of the sequences; 3) Human recoveries as appropriate to reflect procedural changes and operator practices; 4) Initiating events frequencies to reflect current state of knowledge; 5) Data analysis to combine ANO-1 and ANO-2

A*W o t 1CAN059902 Page 15 of 55 failure rates; 6) Anticipated transient without a scram, Internal flood analysis, and i interfacing system LOCA analysis as appropriate to reflect the more current plant configuration.

Each element discussed above was revised, iWantly reviewed, and documented as an ANO calculation. The remits of this work have been summarized in ANO engineering report #95-R-1010-01. This report also documents the CDF for the latest revision of the ANO-1 PSA model (2.01E-05/rx yr).

-In regard to LERF, the containment system work packages were revised as discussed above. An analysis for calculating the LERF value for the updated model has been developed, but has not been completed at this time. Therefore, a new LERF value for the latest model has not been determined.

Entergy Operations Response to 5c It is recognized that the treatment of human errors has a very significant impact on risk assessments that are performed with the PSA. The treatment of human errors, however, is not an exact science but more of an art, since a realistic approach to modeling human actions does not currently exist. As such, the treatment of human error quantification is a weakness in all IPEs. Thus, the statement tint human error evaluation was a " weakness" cannot be considered plant tpecific.

However, in response to the concern regarding the sensitivity of these issues in relation to the RI-ISI analysis, we believe the results and conclusions are not sensitive to human reliability analysis (HRA) values for the following reasons:

• Transients (TI), which are judged to be most sensitive to human response assumptions, have a " low consequence CCDP of about IE-6" (see response to Question 6a), but were treated as a " Medium" consequence. two orders of magnitude in margin exist between the conservatively assumed " Medium" category and the "High" category.

• LOCAs (S, M, and L) are all "High" consequence. This classification would place the LOCAs in a conservative classification and would not be affected by any sensitivity analysis pafu ..d on the HRAs

• Only one human action is specifically analyzed in the RI-ISI evaluation. This action, isolation of the break, is discussed in the response to Question 6b.

__e.

Attachment to 1CAN059902 Page 16 of 55 NRC Question No. 6 la order to conclude that there is reasonable assurance that the results and conclusions appropriately awflect the application of the asethodology to the plant, please clarify the following issues.

a. NSD-024, Section 3.3, " Plant IAvel Assumptions," item 4, page 18, identifies that breaks la the low-pressure injection (LPI)/ core flood piping are limited to 9 inches by flow restricting orifices la the vessel mozzles. It also states that the q

ANO-1 IPE did not include a Medium LOCA, but only small and large LOCAs.

Item 5, page 18, indicates that Reference 3 included a medium LOCA designation but the reference was not used and item 5 concludes that "the  :

present analysis is conservative." Item 6, page 18, states that, "different size LOCAs should be considered." Finally, the second paragraph under Section 4.3,

" Impact Group Assessment," page 27, refers % a table os page 3.1-13 in the IPE where medium LOCA success criteria are gives (although, as stated earlier, not used in the IPE). The text on page 27 and the note (4)(b) on page 36 both apparently use the medium LOCA success criteria from the IPE table on page 4 3.1-13. If success criteria different from those used in the IPE are used, this should be clearly stated and consistently labeled, and all success criteria should be included in Figure 3-1, " Simplified Success Criteria," in the NSD-024 subunitial

b. NSD-024, page 25, credits the operators following the " Reactor Trip Procedure" to close CV-1009. Since the pipe rupture will cause a LOCA, will the operators still follow the Reactor Trip Procedure or will they be following another procedure? Also, item 6, page 18, states that it appean the valve could be closed in time to affect the success criteria but further evaluation could increase confidence. Was this evaluation done? Conversely, the assumption that the letdown isolation valves CV-1213 and CV-1215 do not close automatically following a LOCA, leads to assigning the downstressa segments a High ,

consequence. Do these valves close automatically upon LOCA indication and I are the operators directed by procedure to isolate letdown following a LOCA? )

c. Are the lines between the pressurizer and the safety valves PSV-1000, PSV-1001, and PSV-1002 included in the analysis?

Eatergy Operations Response to 6a There are two subjects that require clarification: (1) the treatment and importance of Medium LOCA and its success criteria, and (2) the analysis of LPI paths with flow restrictors in the reactor vessel nozzles.

I Attachment to-  !

ICAN059902 Page 17 of 55 (1) With regard to Medium LOCA treatment and importance-e Only Small (SLOCA) and Large LOCA success criteria (NSD-024 Figure 3-1) and associated PRA results (NSD-024 Tables 2-1, 3-1, and 3-2) were used in the analysis.

e The ANO-1 PRA (IPE submittal) did not include medium LOCA (MLOCA); it was included within the SLOCA analysis (<4 inch nominal pipe diameter break).

• Subsequent to the IPE submittal, Reference 3 to NSD-024 identified the intent to add MLOCA to the PRA. In the NSD-024 evaluation, this was not a concern, since the success criteria was judged to be essentially the same as for SLOCA and the break size was within the IPE SLOCA definition.

e The IPE has been updated and is referred to as the ANO-1 PSA (95-R-1010-01, Rev 1). The following table summarizes these latest results which include MLOCA:

)

Initiating Event CDF CCDP  ;

Frequency (IEF)

Initiating Event

)

(events /yr) (events /yr) (CDF/IEF) i S - Small LOCA (3/8"-1.9") 5.00E-03 1.13E-05 2.26E-03 TI - Reactor /rurbine Trip 1.83 1.83E-06 1.00E-06 M - Medium LOCA (1.9"-4.3") 1.00E-03 2.13E-06 2.13E-03 L - Large LOCA (>4.3") 1.00E-04 3.66E-07 3.66E-03 Notes: (1) NSD-024 assigns nominal pipe diameter >4 inch to LLOCA.

(2) The above CCDP for SLOCA and LLOCA is less than the IPE and NSD-024 by more than a factor of 2.

e The following conclusion can be derived from the above results and a review of the PSA:

Treating MLOCA as SLOCA in NSD-024 (and the IPE) is reasonable and does not impact the results. The success criteria relative to high pressure injection (HPI) and LPI mitigating equipment is essentially identical for MLOCA and SLOCA; the time window for providing primary system inventory is longer for SLOCA and depends on secondary heat removal.

(2) With regard to the LPI paths:

'e There is a typographical error in NSD-024 Table 4-1 for the entry "CCA-6-14,12 & 8

- upstream of check valve DH-14A or B." SLOCA should be changed to LLOCA for

.the potential (PLOCA) case; CCDP is based on LLOCA. Also, "<8" should have been "8" in this table because the intent was to round off to whole numbers (e.g., 7.5

=8).

Attachment to ICAN059902 Page 18 of 55 1

e I The success criteria discussion in NSD-024 Section 4.3 and Table 4-1 note 4(b) is not referring to MLOCA success criteria. It is referring to the lower end of the LLOCA )

break size (4 to 10 inches). NSD-024 (Section 4.2, top of Page 23) assigns >4 inch j nominal pipe size to LLOCA (PRA uses >4.3 inches). l e The purpose of discussing 4 to 10 inch piping'and the 9 inch flow restrictors is to point out that these pipe breaks are less severe than for the high end of the LLOCA l break (e.g., cold and hot leg piping) and to provide additional qualitative justification for using the LLOCA CCDP (7.5E-3).

I Entergy Operations Response to 6b l

At Step 26, reactor trip procedure (1202.01, Rev 27, April 1999) has operators close both )

CV-1008 and CV-1009 if reactor pressure is less than 1700 psig. There is also an exit to j the emergency safety actuation system (ESAS) procedure (1202.10, Rev 4, April 1999),

for the case where ESAS actuation alarms are annunciated. However, Step 5 of 1202.10 ]

also requires closure of CV-1008 and CV-1009.

l Two different leak sizes were investigated by plant simulator mns: I e A 1.0 square inch break,' which represents the approximate maximum leak size for 1

- this path (based on the spray nozzle). CV-1008 is a control valve and will be closed based on pressurizer pressure

• A 0.72 square inch break to investigate the sensitivity of a smaller leak.

The results of both nms were very similar, with the only difference being a difference in the timing ofevents.

For the 0.72 square inch run, the time from event initiation to ESAS actuation (at 1550 psig RCS pressure) was 3 minutes and 10 seconds, with ESAS actuation approximately 20 seconds aRer automatic reactor trip (at 1800 psig) on low RCS pressure.

For the 1.0 square inch run, the time from event initiation to ESAS actuation was 2 minutes end 5 seconds, with ESAS actuation again occurring approximately 20 seconds after autoiaatic reactor trip on low RCS pressure These results indicate that, even with operator intervention to manually trip the reactor and initiate HPI, it would be improbable that the ESAS actuation (automatic or manual) would be avoided due to the rapid depressurization associated with a steam space leak. ,

In no event will normal makeup be able to maintain pressure for the time required to l progress through the reactor trip emergency operating procedure to Step 26 which would close the spray valves. An ESAS actuation (manual or automatic) requires that the  !

reactor trip procedure be exited and the ESAS procedure be entered. It was estimated to take approximately one and a half minutes (from the time the procedure was entered) to reach Step 5 in the ESAS procedure which would direct closure of the spray valves. ,

i

Attachment to ICAN059902 Page 19 of 55 The above simulator runs demonstrate a high likelihood that operators will ensure that CV-1009 and CV-1008 are both closed within minutes aRer pipe break. With successful isolation, the event turns into a transient.

Letdown valves CV-1213 and CV-1215 do not close automatically on a LOCA signal; other valves downstream perform the isolation function. The operators may diagnose and close CV-1213 and CV-1215, but there is no explicit procedural direction to do so. The only procedural direction to close these valves is if the nuclear intermediate cooling water (ICW) process monitor (radiation alarm) is annunciated (Step 7 of 1202.010, ESAS).

Thus, isolation of CV-1213 and CV-1215 would occur only if there was a heat exchanger leak to the ICW system.

Entergy Operations Response to 6e As described in NSD-024, Table 1-1, these valves are bolted directly to the pressurizer, i.e., there are no Category B-J welds.

I

Attachment to ICAN059902 Page 20 of 55 NRC Question No. 7 The treatssent of breaks la the LPI injection lines upstream of DH-14A/B is not clear; please clarify the foBowing issues:

a. NSD424, page 28, states that ruptures in these lines do not fall both LPI tralas through Row diversion since there are now orifices " upstream of the break,"

meaning upstream of DH-14A/Bf Flow orinces upstream of the break limit the size of the resulting IhCA but not the LPI/ core flood now. The figure on page C-17 la the IPE indicates that it is not possible to isolate one LPI injection line from the other between DH-13A/13B/17/A8 and DH14A/B. Breaks downstream of DH14A/B am High so the assumption is immaterial However, for a break upstream of DH 14A/B, why aren't both LPI traias and one core Good traia mot lost out the break? ,

b. Why is failum of the segment upstream of DH-14A/B labeled a Small LOCA la Table 4-1 when, even with the orifice, it should be at least a Medium LOCA?

4

c. In addition, the CCDP column for this entry uses 1x10 for passive failure of check valve DH-14A or B, but note 5 to the table indicates this value is 1.4x104 ,

d which if used, would result in a CCDP of 1.1x10 even if the LOCA CCDP of 4 4 7.5x10 from Table 2-1 is used rather than 8x10 . Therefore, the CCDP would result la a High consequence ranking rather than the Medium ranking as noted. Please explain why your plant specific data was not used for this scenariof l

Entergy Operations Response to 7a i There is a cavitating venturi in each of 4 LPI lines (two 12" primary lines and two 8" cross-connect lines) in the reactor building and upstream of the check valves and tie-in to ,

the core flood lines. From the LPI Upper Level Document (ULD-1-SYS-04, Rev. 2, )

effective 2/18/98):

The cavitating wnturis allow adequateflow owr a wide range ofRCSpressures andprennt decay heat (DH) pum runout without manual valw throttling. They also limitflowfrom a rmtured DH or coreflood (CF) injection line. ..... At zero RCS pressure, the broken line and unbroken line are both calculated to pass

+x admately 1800 GPMfor single pum operation.

The ANO-1 SAR (Section 6.1.3.2) states the following for one pump operation at a RCS pressure of approximately 100 psig:

(1) 3,000 gpm ofbomted water is injected into the reactor wsselfor core cooling in the ennt ofa large reactor coolant line ryture, and

Attachment to 1CAN059902 -

Page 21 of 55 (2) Apprarimately 1,000 Rpm ofborated water is injectedinto the reactor vesselfor core coolig in the ennt of a corefloodig line break. Dee cross-connection ....... and associated cavitatig venturis also insure that no greater than 3820 gpmflow will occur through a decay heat removalpump.

Thus, plant design ensures that the reactor core is protected for DH line breaks; 1 of 2 LPI pump trains with 1 of 2 injection paths provide successful injection.

Entergy Operations Response to 7b This is a typographical error, SLOCA should be changed to LLOCA.

Entergy Operations Response to 7c

- Agreed. If 1.4E-2 and 7.5E-3. were used in the analysis, the result would be 1.05E-4.

The calculations in Table 4-1 were intended to be order of magnitude estimates. Either 4

1.05 or 1.1 are considered close enough to 10 to not impact the conclusions. Also, the d

10 estimate is conservative because the latest CCDP from the ANO-1 PRA (see Response to Question 6a) is 3.7x10 versus 7.5x10. In addition, a check valve rupture probability of 1.4x10 is judged conservative (e.g., using the ANO-1 failure mode to spuriously open is judged conservative for these valves when compared with the mpture failure mode from other data sources).

In summary, it is our judgment that PLOCA scenarios are clearly a " Medium" consequence relative to unisolable LOCAs; additional detailed analysis to prove the point was not deemed necessary, i

l

Attachment to 1CAN059902 Page 22 0f 55 NRC Question No. 8 To fully address LERF considerations, use of the generic containment failure probability given a core damage of 0.1 must be shown to be applicable to the core damage scenarios caused by these breaks. Please indicate what conditional containment failure probability is estimated for the types of core damage scenarios laitiated by the ruptures of segments classified as Medium, and compare these to those scenarios which cause the estimated 2.8 percent of large enriy release following core damage.

Entergy Operations Response Reference 2 to NSD-024 (IPE Tables 4.3-4, 4.3-5, 4.6-2 and 4.6-3) provides the necessary data to estimate the frequency of early containment failure, as well as large and large-early failure. This data was utilized to estimate the conditional probability of early containment failure for SLOCA and LLOCA which is all that is necessary to address

" Medium" consequence in NSD-024, Table 4-1:

Initiator Conditional Probability of Early Containment Failure Given CDF SLOCA 0.084 LLOCA 0.038 The above values are conservative because the probability of LERF is a subset of the above probabilities. Therefore, the above values clearly demonstrate that a conditional LERF value is less than 0.1.

l I

i i

I

. i Attachment to ICAN059902 Page 23 of 55 NRC Question No. 9 SRP 3.9.8 requires that the licensee determine whether any risk increases would result from impleasentation of the proposed RI-ISI program, and that cu.aulative j effects are small and do not exceed NRC safety goals. In onier to conclude that the changes la CDF and LERF are, with reasonable assurance, actual risk decreases or at most, a negligible increase, please include a discussion of the impact of the changed inspection program on these metrics. If a qualitative discussion based on simple and straight forward assumptions does not cleariy characterize the change in risk, a quantitative bounding estimate or most detailed quantitative evaluation should be provided.

Enter 3y Operations Response The summary of the number ofinspections in both the current Section XI Program and in the proposed RI-ISI Program are shown in two of the tables which are used later in the quantitative evaluation (Table 9-1, per system and risk category, and Table 9-3, total per risk category). These tables suggest, even without quantitative evaluation, that the application of the N560 Code Case to the ANO-1 plant is likely to result in risk improvement. This conclusion is based on the following observations:

• There is an increase of 8 inspections in the High Risk Category e There is a small reduction in the total number of volumetric inspections (11) e All reductions are in risk categories where degradation mechanisms are not present  !

(Risk Categories 4 and 6)

Evaluating actual numerical changes in risk due to changes in the inspection program is a difficult task due to several factors. These factors include the inherent uncertainty in passive component reliability prediction, and the need to predict the changes in piping reliability due to changes in the inspection program.

Consistent with the ANO-2 application, two approaches have been used to estimate changes in CDF and LERF for ANO-1 in order to demonstrate compliance with RG 1.174. The application of these approaches is presented below.

Bounding Estimate of Removed Locations One simplified way to demonstrate that =cc*ptable risk impacts will not occur from implementation of the proposed RI-ISI program is to estimate the risk impact only due to removing locations from the inspection program (without crediting added locations).

This was done by using the estimates of pipe rupture frequencies for locations proposed for removal from the inspection program, and taking the product of these frequencies with the corresponding estimates of CCDP (see bounding values in Table 9-2). The results of this assessment for CDF are summarized in Table 9-1, which shows that the

Attachment to ICAN059902 Page 24 of 55 4

totz.! bounding change in CDF is less than 10 per year. The change in LERF due to removing inspection locations, if a similar approach is used, is expected to be less than 10* per year. These estimates are indicative of negligible contributions to risk from eliminating inspections and are consistent with the criteria of RG 1.174.

Estimate of the Net Change in Risk due to Inspection Enhancements Another simplified approach to estimate the net change in risk due to both positive and negative influences of inspection program changes was first developed for the pilot application for Vermont Yankee This same approach was also applied to ANO-2.

Bounding estimates for the different CCDP ranges and the pressure boundary failure (PBF) likelihood ranges are given in Table 9-2. As shown in Table 9-2, the highest CCDP in the ANO-1 RI-ISI analysis is 8x10-3, which corresponds to LLOCA events.

The likelihood of PBF is a function of the presence of different degradation mechanisms and/or the susceptibility of these locations to degradation mechanisms. The failure potential rank is based on the relative failure likelihood instead of an absolute number.

The likelihood of PBF for a piping location with no degradation mechanism present is 4

given as xo. and is expected to have a value close to 10 per location year.

Assuming that perfonning an inspection on a specific location will lower the risk at this location, with the likelihood proportional to the probability of detection (POD), the difference in risk can be estimated as given in Table 9-3. In general, this is a conservative assumption, in view of the fact that there are several factors that are impossible to inspect for, and because inspections are generally performed only once within a ten-year period, consistent with the current ASME Section XI Code.

In Table 9-3, three (PODS) are introduced. They are defined below:

POD.= Probability of detecting an existing flaw on the location not exposed to a known degradation mechanism (Assumed POD. = 0.5)

PODn = Probability of detecting an existing flaw on the location exposed to thermal fatigue, by using current Section XI inspection techniques. ,

(Assumed PODn = 0.3) {

l PODr2 = Probability of detecting an existing flaw on the location exposed to thermal

]

fatigue, by using improved EPRI RI-ISI inspection techniques.

(Assumed PODr2= 0.9) l 1

A** e h - to 1CAN059902 ,

Page 25 of 55 As can be seen from Table 9-3, given the above assumptions, the reduction in risk resulting from implementation of the RI-ISI program is:

ARcor = -3

• ox ~-3x10# /yr In general, containment performance for pressurized water reactors (PWRs) has shown a conditional LERF of 0.1. Conditional LERF (CLERF) is not expected to be affected by the reduction in inspections in the Class 1 piping. Therefore, the bounding value of 0.1 is used for CLERF. As such, the ARuar would be expected to be an order of magnitude lower than the ARcor:

ARuar ~-3x10# /yr The above evaluation shows that implementation of the proposed ANO-1 RI-ISI Program will result in a small improvement in plant safety, 1

l .

- 1 Attachment to ICAN059902 Page 26 of 55 i Table 9-1 Bounding Estimate of Risk Impact Due to Removed Inspection Locations L RI-Mt l dhkh. 'C; .. =-- Danese t A Inspessed 9p CDFIncswese Du to t

w;, m 4, m L. 2.. ,; ,

Systeens Caengary t Bank';i / Adechanisons L Carrent RI-158 ' laspections Roanoved IsEpectless RCS 2 High 17,PWSCC 0 2 +2 Improvement

]

TF 5 10 +5 Improvement PWSCC 0 1 +1 Improvement 4 High None 18 7 11 9610 5 Medium TF 0 0 0 No Change 4

6 Medium None 0 0 0 No Change l

Make-up 2 High 17 8 9 +1 Improvement (

& Purific- PWSCC 0 0 0 No Change ation 4 High None 6 2 -4 3L10

)

(MUAP) 5 Medium 17,10 SCC 0 1 +1 Improvement TF 0 1 +1 Improvement

)

6 Medium None 0 0 0 No Change l DH 2 High TF 7 5 2 259 PWSCC 0 1 +1 Improvement 4 High None 3 1 -2 2L10 6 Medium None 4 0 -4 4512 All High Risk Region Various 20 28 +8 Improvement RI-ISI Medium Risk Regma Various 27 12 -15 169 Systems Low Risk Regan None 4 0 -4 4512 l

l

l Attachment to 1CAN059902 Page 27 of 55 l

TABLE 9-2 Bounding Estimates for Risk Matrix CONSEQUENCE RANK c LIKELIHOOD OF PBF.c h, ,

, ,' >, [ Rank ( , i CCDPh , ,

. Rank l >

5 Relative PBF+

e , s11 NUppefBesid)l iLikelibssd1',

High SE-3 High 200xo Medium 1.0E-4 Medium 20xo Low 1.0E-6 Low xo TABLE 9-3 Estimate of Change in Risk

[RIAk(CDP)j" , jNorofj ' j No. of::

LRiski hN l tio ,

% Risk (CDF)lj '

Citsgdh " ((i/YR] * ' [ Inspections lasNtioAs ([l/YRjl 1 N/A N/A N/A N/A 2 0.16xo 20 28 (20PODn - 28PODn)

• 0.16xo 3 N/A N/A N/A N/A 4 8E-3xo 27 10 17 POD.

• 8E-3xo _

5 2E-3xo 0 2 -2PODn

• 2E-3xo 6 IE-4xo 4 0 4 POD.

• IE-4xo 7 N/A N/A N/A N/A 3.2PODn

• xo-4.484PODn

• xo

+ 0.1364 POD.

• xo= -3xo

l l

Attachment to 1CAN059902 Page 28 of 55 NRC Question No.10 In the original ISI Program Plan, Table 4.1, " Inservice Inspection Summary Table,"

a total of 444 B-J components are identified. Of these 444 elements,29 are socket 3

welds. The RI-ISI program (Table 3-4 ofIR-98-055) list a total of 394 elements, not including socket welds. Describe the discrepancy and how the total of 394 was determined.

1 Entergy Operations Response The original ISI Program Plan referenced in Question 10 has been superceded by program revision 1, dated October 29, 1998, and submitted on December 9,1998 (ICAN129801), in response to a NRC RAI. This program revision is considered to be the appropriate basis for comparison to the risk-informed submittal. The program  ;

revision submittal will hereafter be referred to as the "RAI submittal" in the remainder of this response. A comparison of the RAI submittal and the RI-ISI submittal resulted in the ider.tification of the following differences.

. The RAI submittal identified a total of 448 elements, while the RI-ISI submittal I contains a total of 394 elements (excludes the three socket welds referenced in NRC )

l Question No.10).

e The three socket welds identified in the RI-ISI submittal were identified as butt welds in the RAI submittal.

e The RAI submittal contained one repeated element (22-064).

• The RAI submittal included 52 longitudinal seam welds (Category B9.12) which were not depicted in the RI-ISI submittal.

e The RAI submittal included a total of 21 elements which were not in RI-ISI submittal scope due to size, not a full penetration weld, etc.

• The RAI submittal contained 6 element numbers that no longer exist due to plant modification / renumbering and therefore are not reflected in' the RI-ISI submittal.

These modifications were primarily in the reactor coolant drain system with the effect of eliminating socket weld connections.

• The RAI submittal contained 2 element numbers that had been inappropriately assigned (not an element).

• A total of 31 items that were contained in the RI-ISI submittal were not listed in the '

RAI submittal. These items were new elements / element numbers generally created as a result of plant modifications (e.g., the reactor coolant drain system) or other items i

E- ,

Attachment to ICAN059902 Page 29 of 55 identified for inclusion as a result of the as-built plant review performed in creating the RI-ISI element list.

Summarizing these element count deltas:

RAI Submittal RI-ISI Submittal 448 Total Count 394 Total Count

-3 Socket Welds -

-1 Duplicate -

-52 Long. Welds -

-21 Not in Scope -

-6 Modification -

-2 No element -

+31 Additions in RI-ISI -

394 Elements- 394 Elements A listing of the elements not appearing in both submittals is provided in the tables on the following pages.

Conclusion In summary, the RI-ISI submittal reflects the current as-built configuration of the plant and has been reconciled with the RAI submittal as demonstrated above.

4 l

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o Attachment to 1CAN059902 Page 30 of 55 Elements in the RAI submittal. but not in the RI-ISI submittal; Longitudinal Welds ll 1 ll 06-006 1 CCA-1 L Long. Welds not independently considered for separate selection ll l 2 l 06-007 l CCA-1 l Long. Welds not independently considered for separate selection l l 3 l 06411 l CCA-1 l Long. Welds not independently considered for separate selection l l 4 l 06412 l CCA-1 l Imag. Welds not independently considered for separate selection l l 5 l 07-004 l- CCA-1 l Long. Welds not independently considered for separate selection l l 6 l 07-005 l CCA-1 l Long. Welds not independently considered for separate selection l l 7 l 07-009 l CCA-1 l Long. Welds not independently considered for separate selection l

- l 8 l 07-010 l CCA-1 l Long. Welds not independently considered for separate selection l l 9 l 07-014 l CCA-1 l Long. Welds not independently considered for separate selection I l_10 l 07-015 l CCA-1 l I4ng. Welds not independently considered for separate selection l l 11 l 08-006 l CCA-1 l Long. Welds not independently considered for separate selection l l 12 l 08-007 l CCA-1 l Long. Welds not independently considered for separate selection l l 13 l 08-011 l CCA-1 l Long. Welds not independently considered for separate selection l l 14 l 08412 l CCA-1 l Imag. Welds not independently considered for separate selection l l 15 l 09-003 l CCA-1 l Long. Welds not independently considered for separate selection l l 16 l 09-004 l CCA-1 l 14ng. Welds not independently considered for separate selection l l 17 l 09-008 l CCA-1 l Long. Welds not independently considered for separate selection l l 18 l 09-009 l CCA-1 l Iang. Welds not independently considered for separate selection l l 19 l 09-012 l CCA-1 l Long. Welds not independently considered for separate selection l l 20 l 09-013 l CCA-1 l 14ns. Welds not independently considered for separate selection l l 21 l 10-006 l CCA-1 l Long. Welds not independently considered for separate selection l l 22 l 10 007 l CCA-1 l Long. Welds not independently considered for separate selection l l 23 l 19411 l CCA-1 l Long. Welds not independently considered for separate selection l l 24 l 16-012 l CCA-1 l Long. Welds not independently considered for separate selection l l 25 l 11-003 l CCA-1 l Long. Welds not independently considered for separate selection l l 26 l 11-004 l CCA-1 l Iang. Welds not independently considered for separate selection l l 27 l 11-008 l CCA-1 l 14ms. Welds not independently considered for separate selection l l 28 l 11-009 l CCA-1 l 14ag. Welds not independently considered for separate seiection l l

l Long. Welds not independently considered for separate selection l l 29 l 11-012 l CCA-1 l 301 11-013 l CCA-1 l Imag. Welds not independently considered for separate selection l

• l 31 l 12-006 l CCA-1 l Long. Welds not independently considered for separate selection l l 32 l 12-007 l CCA-1 l 14ag. Welds not independently considered for separate selection l l 33 l 12-011 l CCA-1 l Long. Welds not independently considered for separate selection l l 34 l 12-012 l CCA-1 l Long. Welds not independently considered for separate selection l l 35 l 13-003 l CCA-1 l Long. Welds not independently considered for separate selection l l 36 l 13-004 l CCA-1 l Long. Welds not independently considered for separate selection l l 37 l 13-008 l CCA-1 l Long. Welds not independently considered for separate selection l l

Attachment to ICAN059902 Page 31 of 55 B9.12 Elements (cont'd) il 38 0 13-009 1 CCA-1 il Long. Welds not independently considered for separate selection l-l 39 l 13-012 l CCA-1 l Long. Welds not independently considered for separate selection l l 40 l 13-013 l CCA-1 l Long. Welds not independently considered for separate selection l  ;

l 41 l 14-003 l CCA-1 l Long. Welds not independently considered for separate selection l l 42 l 14-004 l CCA-1 l Long. Welds not independently considered for separate selection l l 43 l 14-007 l CCA-1 l Long. Welds not independently considered for separate selection l l 44 l 14-008 l CCA-1 l Long. Welds not independently considered for separate selection l j l 45 l 14-024 l CCA-1 l Long. Welds not independently considered for separate selection l l 46 l 14-025 l CCA-1 l Long. Welds not independently considered for separate selection l l 47 l 15-003 l CCA-1 l Long. Welds not independently considered for separate selection l l 48 l 15-004 l CCA-1 l Long. Welds not independently considered for separate selection l l 49 l 15-007 l CCA-1 l Long. Welds not independently considered for separate selection l l 50 l 15-008 l CCA-1 l Long. Welds not independently considered for separate selection l l 51 l 15-023 l CCA-1 l Long. Welds not independently considered for separate selection l l 52 l 15-024 l CCA-1 l Long. Welds not independently considered for separate selection l Not in Sco se 1 17-002 ll CCA-8-12 ll 1"- Not in Scope ll 2 17-005 l CCA-8-12 l 1"- Not in Scope l 3 20-038 l CCA-5 l 1"- Not in Scope l 4 21-056 l CCA-5 l 1"- Not in Scope l 5 21-061 l CCA-5 l Tack weld for relocation of Code Stamp ID Band l l 6 21-102 l CCA-5 l 1"- Not in Scope l 7 21-103 l CCA-5 l  %"- Not in Scope I l

8 22-036 l CCA-5 l 1"- Not in Scope l 9 22-059 l CCA-5 l  %"- Not in Scope l 10 22-063 l CCA-5 l 1"- Not in Scope l 11 23-025 l CCB-5 l Class 2 - not in scope l 12 23-026 l CCB-5 l Class 2 - not in scope l ,

13 23-027 l CCB-5 l Class 2 - not in scope I 14 23-028 l CCB-5 l Class 2 - not in scope l 15 23-054 l CCA-5 l  %"- Not in Scope l 16 23-058 l CCA-5 l 1"- Not in Scope l 17 23-108 l CCA-5 l Arc strike weld repair l 18 23-118 l CCA-5 l  %"- Not in Scope l 19 24-028 l CCA-3 l 1"- Not in Scope l 20 24-037 l CCA-3 l 1"- Not in Scope l 21 24-054 l CCA-3 l 1"- Not in Scope l

Attachment to 1CAN059902 Page 32 of 55 Plant Modifications 01 ll 24-045 ll CCA-13 ll No longer exists due to plant modification / renumbering li l 2 l 24-046 l CCA-13 l No longer exists due to plant modification / renumbering l l 3 l 24-047 l CCA-13 l No longer exists due to plant modification / renumbering l l 4 l 24-048 l CCA-13 l No longer exists due to plant modification / renumbering l l 5 l 24-049 l CCA-13 l No longer exists due to plant modification / renumbering l l 6 l 24-050 l CCA-13 l No longer exists due to plant modification / renumbering l Non-existent Locations j1 ll 24-066 ll CCA-3 11 Element doesn't exist (# assigned to non-existent location) ll l 2 l 25-020 l CCA-13 l Element doesn't exist (# assigned to non-existent location) l i

1 1 ll 06-002 1 CCA-1 ll Identified in as-built review ll l 2 l 06-009 l CCA-1 l Identified in as-buut review l l 3 l 07-003 l CCA-1 l Identified in as-built review I l 4 l 08-002 l CCA-1 l Identified in as-buut review l l 5 l 08-009 l CCA-1 l Identified in as-built review l l 6 l 09-002 l CCA-1 l Identified in as-buut review l l 7 l 10 002 l CCA-1 l Identified in as-built review l l 8 l 11-002 l CCA-1 l Identified in as-built review l l 9 l 12-002 l CCA-1 l Identified in as-built review l l 10 l 12-009 l CCA-1 l Identified in as-buut review l l 11 l 13-002 l CCA-1 l Identified in as-built review l Identified in as-built review l 12 l 16-001 l CCA-2-10 l l l 13 l 17-017 l CCA-8-12 l Identified in as-buut review l l 14 l 18-001A l CCA-4 l Identified in as-built review l l 15 l 19-026 l CCA-6-14 l Identified in as-built review l l 16 l 19-033 l CCA-6-12 l Identified in as-built review l l 17 l 24-067 l CCA-13 l Identified in as-built review l l 18 l 25-001 l CCA-13 l Identified in as-built review l l 19 l 25-011 l CCA-13 l Identified in as-built review l l 20 l 25-029 l CCA-13 l Identified in as-built review l l 21 l 25-030 l CCA-13 l New element number as result of plant modification l l 22 l 25-037 l CCA-13 l New element number as result of plant modification l l 23 l 25-038 l CCA-13 l New element number as result of plant modification J

l 24 l 25-039 l CCA-13 l New element number as result of plant modification l l 25 l 25-040 l CCA-13 l New element number as result of plant modification l

Attachment to l ICAN059902 Page 33 of 55 Elements in RI-ISI submittal not included in the RAI submittal:

26 25-041 CCA-13 New element number as result of plant modification 27 25 042 CCA-13 New element number as result of plant modincation 28 25-043 CCA-13 New element number as result of plant modiGcation 29 25-044 CCA-13 New element number as result of plant modification '

30 25-045 CCA-13 New element number as result of plant modification 31 25-046 CCA-13 New element number as result of plant modincation i

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AHa4= ant to ICAN059902 Page 34 of 55 l

NRC Question No.11 Qualification of mondestructive examination (NDE) systems (personnel, procedure, and equipment) is sa important element of the RI-ISI program. The relisbuity of  ;

emannimations must be established to achieve the desired confidence levels for the risk-laformed inspection process. Therefore, the technical basis for the inspection reliability impets used la structural reliability calculations of estimated failure probabilities must be justified. Such a basis can be provided by NDE performance demonstration programs. It is maclear how NDE methods, procedures, and personnel will be qualified at ANO-1. With respect to this area provide the  !

foBowing information: I a) Provide a detaBed technical discussion describing how the reliability of NDE i performed in the RI-ISI program wul be qualified. I b) WHI ANO-1 use the licensee's own examination procedures, equipment, and persommel to perform ultrasonic examinations of selected examination volumes?

c) If so, are the procedures, equipment, and personnel qualified la accordance with the requirennents of Appendix VII of the ASME Code,Section XI?

d) If contractors are used to perform these examinations, are the contractors required to use procedures, equipment, and persommel qualified to Appendix VIII of Section XI? ,

Entergy Operations Response In general, the NDE in question are volumetric examinations (i.e., ultrasonic examinations).- Effective January 1,1998, Entergy Operations at Arkansas Nuclear One began using only ultrasonic examinatio. (UT) personnel that meet the qualification requirements of the 1992 Edition of ASME Section XI, Appendix VII, " Qualification of Nondestructive Examination Personnel for Ultrasonic Examination." Additionally, only UT personnel that have successfully completed the practical examinations of the Performance Demonstration Initiative (PDI) administered by the EPRI NDE Center, for

. the specific application areas, (e.g., carbon steel, austenitic steel, intergranular stress corrosion cracking, etc.) are used for ASME Section XI ultrasonic examinations. This practical performance demonstration is a comprehensive qualification for UT personnel.

It is used as the practical examination required by Appendix VII and to meet contractor screening requirements for UT at Entergy Operations' nuclear sites. The PDI assures that UT personnel have experienced a variety of examination problems and the associated difficulties of flaw detection and discrimination. This test enhances the skills of the UT personnel and provides a measurable scale of the reliability of those skills.

Entergy Operations will continue to use station procedures, equipment, and personnel to perform the required ultrasonic examinations. Personnel are qualified by virtue of successful completion of the PDI UT qualification, which is recognized in the site written practice as the required Appendix VII practical demonstration test. Equipment becomes qualified once succeufully demonstrated at the PDI. ANO procedures use the

Attachment to 1CAN059902 i Page 35 of 55 l methodology qualified in the generic procedures at PDI, however, the site specific station {

procedures have not been qualified to Appendix VIII. Contractors used at ANO are I required to meet the same requirements as station personnel and use procedures provided  !

by ANO. Any contractor supplied equipment meets the qualification requirements established at PDI.

)

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Attachment to 1CAN059902 Page 36 of 55 NRC Question No.12 The licensee has selected 40 B-J elements for examination based upon a consequence evaluation and a degradation mechanism evaluation. Appropriate examination methods should be selected to address degradation mechanisms, pipe sizes, and material of concern. It appears that the only method of examination to be performed on the subject elements is volumetric examinations. Provide the specific NDE methods to be used for each of the subject elements. Provide information describing why the particular NDE method was chosen as well as why other NDE methods were not choses (i.e., surface examinations and visual examinations). 1 I

Entergy Operations Response A listing of the 40 RI-ISI element selections is provided in the response to Question 14.

For the risk category 2 and 5 element selections, an inspection for cause process shall be implemented utilizing examination methods and volumes defined specifically for the degradation mechanism postulated to be active at the inspection location. The EPRI RI-  ;

ISI methodology specifies volumetric examination for the' degradation mechanisms identified in the ANO-1 RI-ISI application (i.e., thermal fatigue, intergranular stress corrosion cracking, and primary water stress corrosion cracking). For risk category 4 inspection locations (i.e., no identified degradation mechanisms), the examination methods and volumes to be used are based upon the requirements for thermal fatigue.

This approach for such locations is consistent with the ongoing revision to Code Case N-560 as well as the ANO-2 and Suiry pilot applications. In all cases, UT will be the volumetric examination method employed.

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Attachment to ICAN059902 Page 37 of 55 NRC Question No.13 The inservice inspection strategy used in the RI-ISI program must define when the i inspections are to be performed. Specified inspection latervals must be consistent I with relevant degradation rates. Inspection intervals should be sufficiently short such that degradation too small to be detected during one inspection does not grow to na unacceptable size before the next inspection is performed. Provide a discussion regarding the inspection intervals contained la the RI-ISI program, include an examination schedule and confirm that the proposed examination frequency will not exceed the current Section XI inspection interval of ten years.

Entergy Operations Response The frequency ofinspection that will be applied at ANO-1 is identical with the 10-year inspection interval required by the current ASME Section XI program supplemented with augmented inspections for specific degradation mechanisms, consistent with all other pilot plant applications. These inspections for ANO-1 will provide an acceptable level of quality and safety. As discussed in other applications and methodology submittals, the inspection schedules for aggressive mechanisms such as flow accelerated corrosion (FAC) are based upon a more frequent examination schedule. For mechanisms with slow growth rates, which are typical of those covered by the ANO-1 submittal, where operating experience has shown that there is no need for augmented inspections, the Section XI inspection interval has been used successfully in the past and will continue to be used to manage these mechanisms. Inspection at an increased frequency is incorporated where deemed appropriate. For example, augmented inspections of the HPI nozzle knuckle regions and thermal sleeves will continue to be conducted every 5*

refueling ouiage as discussed in the response to Question 18.

The ANO-1 RI-ISI program will commence during the first period of the third inspection interval, with examinations allocated in accordance with IWB-2400. The sequence of piping examinations established during the current inspection interval using the risk-informed process will be repeated during each successive inspection interval. An examination schedule, for the RI-ISI program, has not yet been prepared for the third inspection interval pending NRC approval of the ANO-1 pilot application. Once l constructed, however, the risk-informed examination frequency will conform with the i

Section XI inspection interval as discussed above.

. The ANO-1 service history and susceptibility review conducted as part of the RI-ISI evaluation and ongoing industry events reviews assure that ANO-1 and industry trends are being monitored, These reviews ensure that if an unexpected or new mechanism it identified, or a new component is identified as susceptible to an existing degradation mechanism, the RI-ISI program can be updated to reflect that change.

Operating experience is assessed at ANO through Procedure 1010.008, Industiy Events and-Analysis Program This program requires that Entergy Operations review industry

Attachment to ICAN059902 Page 38 of 55 operating experience for its applicability to ANO programs and practices. The purpose of this program is to insure that lessons learned from industry and in-house operating experience are evaluated for applicability to ANO, and when appropriate, translated into corrective actions to minimize the potential for similar events occurring at ANO.

1 l

Attachment to 1CAN059902 Page 39 of 55 NRC Question No.14 Considering that the implementation of the proposed RI program will significantly reduce the number of examinations, limited examinations could have a significant impact on the risk. The licensee has chosen multiple elements that have not received previous inspections. Have the elements been choses such that examination limitations do not exist? In the case that examination limitations are encountered on any elements, specify what alternatives will be used to ensure structural lategrity.

Entergy Operations Response As identified in the attached table, there are 40 RI-ISI selections, 29 of which have been previously inspected by volumetric and/or surface means. Accessibility is an important consideration in the element selection process. As auch, locations were generally selected for examination where the desired coverage is achievable. However, some limitations will not be known until the examination is performed, since some locations will be examined for the first time.

In addition, other considerations can take precedence and dictate the selection of locations where complete examination coverage may not be physically possible. This is especially true for the RI-ISI selections in risk categories 2 and 5. For these risk categories, elements were generally selected for examination on the basis of predicted degradation severity.

The element selection process is based on the consideration of a number of diverse factors. The follow' gmexcerpt from revised TR-112657 describes the factors considered in the element selection process This criteria was employed in the ANO-1 application.

Plant Specyle Service History: Ihe results of the plant specific service history review are a key input in the element selection ymcess. Prior idennpcation of piping cracks or flaws potentially sigmpes the presence of an actin damage mechanism.

P>ndicted Severity of P6stulated Damage Mechanisms: Engineering Judgement should be applied to assess the relatin senrity (e.g., delta temperature or Richardson Numberfor thermalfatigue) of postulated damage mechanisms. An exarrpie of this type ofconsideration is pmvided below.

In the emergency core cooling system irgfection lines of PWRs the piping section immediately systream of.<heprst isolation check vain is considered susceptible to intergranular stress cormston cmcking (IGSCC) assuming a syficiently high temperature and oxygenated water supply. The piping element (pipe-to-valw weld) located nearest the heat source will be subjected to the highest temperature (conduction heating). As such, this location will generally be selected for

Attachment to ICAN059902 Page 40 of 55 examination since it is considend more susceptible than locationsfurther remond fmm the heat source, enn though a pipe-to-mlw wid is inhenntly more dWicult to examine and obtainfull cowrage on than most other coryigurations (e.g., pipe-to-elbow wid). (Authors' Note: In this example, limited coverage of this location will yield far more valuable information than 100% coverage of a less susceptible location.)

Co#mdon / Accessibility of Element to Enable 58ective Examinadon: When possible, elements should be selected such that a complete examination of the nquired wlume can be acconplished. Elements that are physically obstructed (e.g., suppon pipe clanp) should be awided as should element cortigurations that are inherently more d@fcult to examine unless other considerations take precedence (see example abow).

Radiation Exposun: In general, elements should be selected consistent with ALARA principles (10CFR20).

Senss Concentmdon: For risk category 4, element selections should befocused on terminal ends and structural discontinuity locations of high stress and/or high fatigue usage in the absence of any identified damage mechanisms. In these cases, a greater degne offlexibility exists in choosing inspection locations.

Physical Accou to Element: Consideration may be giwn to selecting elements that an readily accessible (e.g., examination can be performedfrom floor or grating without scqfolding) without the needfor additionalplant support.

Element Selecdon Pacess Considerations Factors Consideredin Element Selection Selection Selecdon Pmeest Plus Nenative PlantService History Poor Good Sewrity ofDamage Mechanisms High Low Element Cortfiguration / Accessibility Good Poor Radiation Exposure Low High Stress Concentration High Low Physical Access Readily DiBicult As seen in the above excerpt from TR-ll2657, for risk category 4, element selections are focused on terminal ends and structural discontinuity locations of high stress and/or high fatigue usage in the absence of any identified damage mechanisms. In these cases, a greater degree of flexibility exists in choosir:g inspection locations. As such, if at the time of examination a RI-ISI element selection is found to be obstructed, a more suitable location should be substituted instead.

0 4

Attachment to ICAN059902 Page 41 of 55 Also note that if an existing ASME Section XI inspection location is partially examined and continues to be partially examined in the RI-ISI process, the amount of risk addressed by examination remains the same for that location. If a new RI-ISI inspection i

location is only partially examined, but was not previously required to be examined by Section XI, then the amount of risk addressed by examination of that location is still increased. It is not necessarily true that because examination totals are reduced, that a complete examination must be performed at the RI-ISI selected locations to maintain risk neutrality or improvement in the program.

l The RI-ISI piping element selection criteria requires that sufficient examination coverage be achieved to confirm the absence of the applicable degradation mechanism (s). Given this requirement, the inspection for cause process employed in the RI-ISI approach in general ensures that the selected elements are adequate'v interrogated to detect a ,

degradation mechanism, if present. This ensures a high prouability of detection exists in l spite of potential examination limitations. Expanded exam volumes and inspection techniques tailored to the identified potential degradation mechanism (s) contribute to this high probability ofdetection.

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Attachment to ICAN059902 Page 46 of 55 i

NRC Question No.15 The proposed RI-ISI program contains 40 elements for examination. Considering that evaluation ofindications is required based on ISI findings, describe the acceptance criteria and the examination expansion philosophy that will be utilized.

Enter 2y Operations Response The acceptance criteria for the ANO-1 N-560 program will be in accordance with the guidance provided in Revised EPRI TR-112657, Section 4. This is consistent with the guidance provided in EPRI TR-106706, Section 7, and the existing ASME Section XI program for Class 1 piping at ANO-1. Table 4-1 of Revised EPRI TR-112657 is provided, for information, on the following page. With respect to examination expansion philosophy, please refer to the response to Question le.

1 l

Attachment to 1CAN059902 Page 47 of 55 Table 4-1 Summary of Degradation 4pecific inspection Requirements and Examination Methods

. Degradation M Degradation ' Examination Examination Acceptance! Evaluation 4Mechanimel! lw MechanismL lloquirement Method' Standard '  ; Standard 1 L, i ,MV pubcategory [ Fig? No. ' '

, ,,1 Thermal 4-1 Volumetdc IWB-3514 IWB-3640 fatigue 42 or 4-3 IWB-3650 4-4 Corrosion Chloride crecidng Affected Surface IWB-3514 IWB-3640 cracking (OD) Surface or IWB-3650 Chloride cracking 4-5 Volumeldc IWB-3514 (ID)

Crevice cormsion 4-6 Volumetdc IWB-3514 4-7 PWSCC Visual (VT-2) IWB-3142 IWB-3640 4-8 Volumetdc IWB-3514 4-9 IGSCC 4-10 Volumetdc IWB-3514 IWB-3640 through 4-14 Microbiolo0- Volumetdc IWB-5250(b) Code Case ically-induced or N-480 cormsion 4-15 Visual, VT3 IWB-5250(b) N-480 with (MIC) volume equivalent thickness Erosion- See FAC Volumetric Same as Same as FAC cavitation FAC Flow- 4-16 Volumetdc Plant Plant program accolorated throu0h program )

corrosion 4-22 l

1. The frequency of inspechon for each degradation category is each inspechon interval, except for the existing plant '

inspection programs for IGSCC and FAC, where the frequencies specified in the plant programs are applicable.

2. Volumetric exerninsbons are generally performed using ultrasonics, unless otherwise indicated.

Attachment to 1CAN059902 l Page 48 cf 55 i NRC Question No.16 Typically, licensee's perform periodic updates to ISI programs. Updates typically consider:

• Plant design feature changes, plant procedure changes, and equipment performance changes.

• Leakage, Raws, or indications identified during scheduled RI-ISI examinations.

• Plant and industry failure information.

Describe the intentions of the licensee, conceralag RI-ISI program update frequencies and modifications (if any) to the RI-ISI plan.

Entergy Operations Response The RI-ISI program is a living program requiring feedback of new relevant information to ensure the appropriate identification of risk significant piping locations. As a minimum, the impact of these changes on the risk ranking of piping segments will be reviewed and adjusted on an ASME period basis. Significant changes may require more frequent adjustment based on review of industry information.

1

, ,4 .

Attachment to ICAN059902 Page 49 of 55 NRC Question No.17 I It is understood that the licensee has excluded vessel nozzle to safe-end welds classified as category B F welds from the evaluation. Are there any examination category B-J dissimilar metal welds between combinations of(1) carbon or low alloy steels to high alloy steels, (2) carbon or low alloy steels to high alcket alloys, or (3) high alloy steels to high alckel alloys, that were used in the evaluation process? If so, how many, and how many have been chosen for examination?

Entergy Operations Response The following tables provide a listing of all Category B-J dissimilar metal welds as well as Category B-F dissimilar metal welds. From the total of 22 Category B-J welds, 6 have been selected for inspection. All seven Category B-F welds will continue to be inspected per existing Code requirements.

Cateoory B-J Dissimilar Metal Welds ll Surge Line-Hot Leg End ll 16-001 4 Yes l l Main Pressurizer Spray Line - Cold Leg End l 18-001A l Yes l l Cold Leg Drain (C) l 06-009 l No l I

l Cold Leg Drain (C) l 25-001 l No l l Cold Leg Drain (D) l 08-009 l No l l Cold Leg Drain (D) l 25-011 l Yes l l Cold Leg Drain (B) l 12-009 l No l l Cold Leg Drain (B) l 25-029 l No l l Reactor Coolant Pump Suction (C) l 06-002 l No l l Reactor Coolant Pump Suction (D) l 08-002 l No l l Reactor Coolant Pump Suction (A) l 10-002 l No l l Reactor Coolant Pump Suction (B) l 12-002 l No l l Reactor Coolant Pump Discharge (C) l 07-003 l No l l Reactor Coolant Pump Discharge (D) l 09-002 l No l l Reactor Coolant Pump Discharge (A) l 11-002 l No l j Reactor Coolant Pump Discharge (B) l 13-002 l No l l High Pressure Injection (C) l 20-045 l Yes l l High Pressure Injection (D) l 21-064 l Yes l l High Pressure Injection (A) l 22-071 l No l i l High Pressure Injection (B) l 23-065 l No _J l Letdown Line l 24-066A l No l

. l Decay Heat Drop Line l 17-017 l Yes l

Attachment to 1CAN059902 Page 50 of 55 Cate 'ory B-F Dissimilar Metal Welds k Pressurizer ReliefNozzles ll 05-040 / 05-041/ 05-042 D l Pressurizer Surge Line - Pressurizer End l 16-012A l l Pressurizer Spray Line - Pressurizer End l 18-001 l l Core Flood Lines (T-2A / T-2B) l 01-025 / 01-026 l l

l

, 4

• O

' Attachment to ICAN059902 Page 51 of 55 NRC Question No.18

. In the licensee's original ISI program, many augmented examinations have been listed for performance. It appears that some of the augmented examinations are related to Class 1 elements. However, it is unclear whether any of the augmented examinations are I effected / eliminated by the RI-ISI program. Is it the licensee's latention to continue to perform all of the augmented examinations as described in the originalISI program?

Entergy Operations Response The following italicized section excerpted from Revised EPRI TR-112657 provides guidance for i the integration of augmented inspection programs with the RI-ISI process Following this excerpt, ANO-1 specific issues are addressed, including a table identifying the augmented programs and the RI-ISIimpact.

Licensees have a number ofinspection programs beyond their Section XI programs. Many of these programs are licensee commitments to the NRC, while others are a result of plant specific experiences and good practice initiatives. The philosophy ofintegrating inspection programs outside the scope of Section XI is summarized in Table 6-2 and described as follows:

(Other degradation mechanism discussions omitted - Only thermal fatigue discussion included i for purpose of this response)

Ihermal Fatigue - A number of plant irupection programs include augmented examinations performed in response to NRC Bulletins 88-08, " Thermal Stresses in \

Piping Connected to Reactor Coolant Systems, " and 88-11, "Pressuriser Surge Line 7hermal Strat@ cation," and Information Notice 93-020, " Thermal Fatigue Cracking of Feedwater Piping to Steam Generators." The EPRI RI-ISI process depnes an explicit set ofatributes that must be considered in assessing the potential susceptibility of a location to thermalfatigue. The thermalfatigue concerns ident$ed by these documents were inputs considered in the development of the EPRI degradation mechanism criteria. As such, this concern is explicitly consideredin the opplication of the EPRIRI-ISIprocess. Consequently, the RI-ISI .

1 program wouldspersede these augmented trupections.

At this time only the qforementioned changes to inspectionprograms outside the scope ofSection A7 are considered in the EPRI RI-ISI process. However, the industry and the NRC are continuing work on piping integrity issues (e.g., NUREG-0313, BWRVIP results and operating emerlence). It is EPRI's intent to keep current with industry and NRC interactions. Any changes to aplant's licensing basis as a result of the above will need to be identified as part of thefollow-onplantprocess and the licensee must assure oppropriate notification consistent with l existingplantprocesses. l

rn,v Attachment to ICAN059902 Page 52 ef 55 TABLE 6-2 Integration ofInspection Protrams JSSUE Ing-% into RI-ISIProgram

• NUREG--0313, Rev 2 (IGSCCin BWRs) Limitedto Categwy A welds only NRC Bulletin 88-08, "ThermalStresses in Yes, .specipcally ad&essed by thermal Piping ConnectedtoRCS" fatigue evaluation j NRC Bulletin 88-11, " Pressurizer Surge Yes, .specifcally ad&essed by thermal Line Stratifcation" fatigue evaluation 1

NRCIn/wmation Notice 93-020, " Thermal Yes, .specapcally ad&essed by thermal Fatigue Cracking ofFeedwater Piping to fatigue evaluation Steam Generatws" IEBulletin 79-17, Pipe Cracks in Stqgnant Yes, .specifcally ad&essed by stress Bwated WaterSystems atPWR Plants cwrasion cracking enaluation Service WaterIntegrityProgram (GL 89 Yes, .specapcally adhessed by localized

13) cwrosion enaluation Flow Accelerated Carasion (GL 89-08) Yes, no change to the number, type w {

frequencyofinspection 1

• - Each of these augmented programs may be integrated with the RI-ISI program at the discretion of the licensee. However, regardless of the above, revision to licensing commitments and qppropriate notifcation needs to be conducted by individuallicensees.

The majority of augmented examinations for ANO-1 are unaff'ected by the RI-ISI program because they are related to components outside the scope of the RI-ISI evaluation. The impact of the RI-ISI evaluation on existing augmented inspection programs documented in the original ANO-1 ISI program is as described in the attached tables and notes.

I 1

l l

)

Attachment to ICAN059902 Page 53 of 55 Assessment of RI-ISI Pavaram Impact on ANO-1 Augmented Inspections I ANO Description of Augmented Programs RI-ISI Program Reference Impact 1.4.1 Ultrasonic examinations on the reactor pressure No impact- subject vessel shall be conducted in accordance with U.S. components not in Nuclear Regulatory Commission RG 1.150, Rev. RI-ISI scope.

1, " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations". '

l.4.2 High Energy Line Break (HELB) and Moderate No impact - subject Energy Line Break (MELB) examinations shall be components not in performed in accordance with Upper Level RI-ISI scope.

Document ULD-0-TOP-07, "HELB/ MELB Topical ULD," and Calculation 86D-1005-29, Appendix B.

1.4.3 Visual inspections shall be performed on High Pressure Injection (HPI) support MU-167 and See Note 1.

associated piping in accordance with Arkansas Nuclear One Condition Report CR-1-96-0502.

1.4.4 The examination of stagnant, borated water No impact - subject systems (i.e., Nuclear Regulatory Commission IE piping not in RI-ISI Bulletin 79-17) is addressed in Request Number scope.97-001.

l 1.4.5 Enh-M ultrasonic examinations shall be performed on 17 HPI welds and visual inspection See Note 2.

shall be performed on two segments of HPI piping in accordance with Nuclear Regulatory Commission IE Bulletin 88-08. ,

1.4.6 Surface and volumetric examinations of reactor No impact - subject coolant pump flywheels shall be conducted in components not in accordance with Arkansas Nuclear One, Unit 1 RI-ISI scope.

Technical Specification 4.2.6.

1.4.7 Ult:asonic examinations on one pressurizer upper No impact - subject level tap shall be conducted in accordance with component not in Arkansas Nuclear One Calculation 86E-0074-103 RI-ISI scope.

and letter nos. ANO-92-00507 and ICAN089302. I 1.4.8 Ultrasonic examinations shall be performed on No impact - subject reactor coolant pump shafts and surface components not in examinations shall be performed on reactor RI-ISI scope.

coolant pump shaft covers in accordance with Plant Impact Evaluation (PIE) 87-0082B, each

]

time a pump is disassembled.

)

. . . c.

Attachment to ICAN059902 Page 54 of 55 Assessment of RI-ISI Program Impact on ANO-1 Augamented Inspections (Cont'd)

ANO Description of Augmented Programs RI-ISI Program Reference Impact 1.4.9 Surface examinations and visual inspections shall No impact - subject be performed on emergency feedwater riser welds components not in in accordance with Babcock & Wilcox letter no. RI-ISI scope.

APL-85-349.

1.4.10 Visual VT-2 inspections shall be performed on No impact- subject pressurizer level taps in accordance with Arkansas components not in Nuclear One Condition Report CR-1-90-0853-07 RI-ISI scope.

and letter no. ANO-92-02496.

1.4.11 Surface examinations and VT-3 visual inspections No impact -subject shall be pcrformed on the reactor pressure vessel components not in internals fixture tripod in accordance with RI-ISI scope.

NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants."

1.4.12 Ultrasonic examinations shall be performed on HPI nozzle knuckles and radiography shall be See Note 3.

performed on HPI nozzle thermal sleeves in accordance with Arkansas Nuclear One Condition Reports CR-1-89-0335 and CR-1-89-0508.

1.4.13 Ultrasonic examinations and thickness No impact-subject measurements shall be performed on four main components not in steam moisture and reheater piping welds in RI-ISI scope.

accordance with Arkansas Nuclear One Condition Report CR-1-89-0621.

1.4.14 Visual inspections shall be performed on dry spent No impact - subject fuel storage casks in accordance with Sections components not in 1.3.2 and 1.3.3 of the storage casks' " Certificate of RI-ISI scope.

Compliance" which was issued in accordance with 10CFR72.

1.4.15 Visual inspections shall be performed on reactor No impact- subject coolant pump seal injection piping and supports in components not in accordance with Arkansas Ih: lear One Condition RI-ISI scope. l Report CR-1-90-0514-11. j 1.4.16 Ultrasonic examinations shall be performed on No impact - subject pressurizer relief valve piping in accordance with components not in Arkansas Nuclear One Condition Reports CR RI-ISI scope.

92-0244 and CR-1-91-0131, Item 8. l I

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. , , . 1 Attachment to 1CAN059902 Page 55 Cf 55 Assessment of RI-ISI Program Impact on ANO-1 Angmented Inspections (Cont'd)

ANO Description of Angamented Programs RI-ISI Program Reference Impact 1.4.17 Ultrasonic examinations shall be performed on the pressurizer surge line elbow u discussed in See Note 4.

Nuclear Regulatory Commission letter no.

1CNA029402 (TAC No. M72108) dated February 24,1994.

Notes:

1. Element 23-053 is a RI-ISI selection adjacent to the support of concern Reference Note 4 of Table 2-2, Service History and Susceptibility Review, SIR-98-055. The visual inspections will continue to be performed.

2. Elements in the subject piping were selected for examination in accordance with the RI-ISI process. The issue is exclusively considered and as such the RI-ISI program effectively supercedes the original commitment.

3. The knuckle region of the HPI nozzles will continue to be ultrasonically examined every 5*

refueling outage in acw,ider.cc with the BWOG recommendation. This knuckle region consists of two welds per nozzle (i.e., elbowdsafe end and safe end-to-nozzle). Of the eight total welds, six are RI-ISI selections. These six, plus the other two, will all be examined at the increased frequency discussed above. The radiography of the thermal sleeves will continue and is not impacted by the RI-ISI application.

4. Surge line elbow body examinations included as RI-ISI selections - i.e., the RI-ISI ,

examinations will include elbow body, j l

l 1

,