ML20140H893

Summary of 961105-06 Meeting in Rockville,Md Which Was Held to Gather Info on Proposed Rule & Reg Guide Re SG Intergity

ML20140H893
Person / Time
Issue date:	11/26/1996
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-3036, GL-95-05, GL-95-5, NUDOCS 9705130309
Download: ML20140H893 (14)

Text

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, '. CERTIFIED: Robert Seale December 3, 1996 hbdMd Date Issued: November 26, 1996 ADVISORY COMMITEE ON REACTOR SAFEGUARDS JOINT SUBCOMMITTEE MEETING MINU1ES:

MATERIALS & METALLURGY AND SEVERE ACCIDENTS NOVEMBER 5-6, 1996 ROCKVILLE, MARYLAND The ACRS Joint Subcommittee on Materials & Metallurgy and Severe Accidents met on November 5-6, 1996, at 11545 Rockville Pike, Rockville, Maryland, in Room T-2 B3. The purpose of the meeting was to gather information on the proposed rule and regulatory guide related to steam generator integrity. The entire i meeting was open to public attendance. Mr. Noel Dudley was the cognizant ACRS I staff engineer for this neeting. The meeting was convened at 8:30 a.m. on November 5, 1996, and adjourned at 12:00 noon on November 6, 1996. I ATTENDEES LCM R. Seale, Chairman D. Powers, Member M. Fontana, Chairman W. Shack, Member G. Apostolakis, Member N. Dudley, ACRS Staff T. Kress, Member NRC STAFF B. Sheran, NRR T. Reed, NRR R. Jones, NRR S. Long, NRR ;Dck J. Strosnider, NRR R. Palla,.NRR J. Hayes, NRR C. Tinkler, RES E. Murphy, NRR J. Gorman, NRR Consultant pl l J. Donoghue, NRR INDUSTRY I M. Tuckman, Duke Power Company T. Pitterle, Westinghouse C. Welty, EPRI J. Smith, Rochester Gas & Electic

0. Steininger, EPRI R. Pearson, Northern States Power '

There were no written comments or requests for time to make oral statements received from members of the public. An attendance list of members of the NRC staff and public is available in the ACRS office files.

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Minutes: Materials & Metallurgy and Severe Accidents

! November 5-6, 1996 I l t

INTRODUCTION:

j Dr. Robert Seale, Chairman of the Joint Subcommittee on Materials & Metallurgy and Severe Accidents, explained that the purpose of the meeting was to gather information concerning the technical approach used in developing the proposed

risk-informed, performance-based rule and regulatory guide associated with )

" steam generator tube integrity. He sumarized the information gathered during the June 3-4, 1996 Joint Subcommittee meeting, and noted that Subcomittee

, Members should identify items for discussion at the November 7,1996 ACRS meeting.

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Dr. Seale noted that Dr. William Shack has a conflict of interest concerning

! the steam generator rule and would not participate in Subcomittee j deliberations.

i STAFF INTRODUCTION: Dr. Brian Sheron, NRR I

Dr. Sheron stated that the staff intended to explain the technical basis for the proposed steam generator integrity rulemaking package, and hoped to i receive ACRS endorsement of the technical basis for going forward with the proposed rule. Dr. Sheron also requested ACRS endorsement for the staff to issue the proposed rule for public coment. He outlined the past ACRS reviews of the rulemaking effort and presented a tentative schedule for future ACRS reviews. He described the types and extent of staff interactions with representatives of the Nuclear Energy Institute (NEI) and industry concerning the rule and regulatory guide. Dr. Sheron stated that two significant policy issues were how to address severe accident risk within the regulatory framework and how much flexibility should be afforded the industry. Dr.

Sheron and Dr. Seale discussed how the inspection staff would be trained to i verify implementation of the proposed rule.

INDUSTRY INTRODUCTION: Mr. Mike Tuckman, Duke Power Company Mr. Tuckman introduced himself as the Vice Chairman of the Electrical Power 3 Research Institute (EPRI) steam generator management program and the Chairman I of the NEI steam generator working group. He sumarized how the industry l supported the development of the degradation specific management concept and the voltage-based alternate repair criteria. Mr. Tuckman noted that EPRI had l conducted assessments of some licensee steam generator tube inspection  ;

programs and that the Institute for Nuclear Power Operations had agreed to i continue the assessments. '

The industry, staff, and Subcomittee Members discussed the following issues:

• whether the NRC should review and approve utility methodologies; e the level of detail in the proposed regulatory guide; e severe accident analyzes becoming part of the licensing basis;

Min'utes: Materials & Metallurgy and Severe Accidents November 5-6, 1996

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the meaning of risk-based, risk-informed, and performance-based; and the adequacy of defence-in-depth requirements and risk analyses to assure the containment function of the steam generator tubes.

NRC STAFF PRESENTATION:

Proposed Steam Generator Rule: Mr. Jack Strosnider, NRR, stated that the i steam generator rule was developed to be performance-base, risk-informed, adaptable, and enforceable. He noted that the rule was designed to provide an incentive for utilities to improve tube inspection technology. He explained

the intent, objectives, and requirements of the rule. Mr. Strosnider noted

that the rule would be a backfit and would supersede the present regulatory requirement.

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The suff and the Subcommittee Members discussed the relationship of tube I inspectici: analyses to the probability of a tube failure, the perscriptiveness of the rule, the addition requirements for analyzing severe accidents, staff approval of performance criteria, and acceptable calculational methods. They also discussed the relationship between measurable and calculable performance criteria, the flexibility of the rule to deal with new tube degradation mechanisms, and the consistency between the structural performance criteria and the accident leakage criteria. Discussions were also held concerning the i .following issues:

. how the staff will assure that methods or approaches to demonstrate conformance with performance criteria are adequate, and whether the rule is necessary if the staff is required to review industry implementation documents and condition monitoring methodologies.

4 Proposed Steam Generator Regulatory Guide: Mr. Emmett Murphy, NR.1, explained the flow diagram of the proposed regulatory guide program strategy. He described the regulatory guide structural performance criteria, the operational leakage criteria, and the accident-induced leakage criteria. He summarized the proposed guidance associated with inservice inspection i objectives, frequency, sample size, and data acquisition and analysis. Mr.

Murphy explained the proposed guidance for the backward looking condition monitoring and the forward looking operational assessment evaluations. He 1 presented the guidance for tube repair criteria, tube repair methods, acceptance criteria, reporting requirements, and preventive measures.

The staff and the Subcommittee Members discussed developing alternate repair criteria correlations, the uncertainty associated with the probabilities used in the regulatory guide, and the randomness of expected spontaneous tube ruptures. They also discussed the meaning of the mean value for conditional probability of failure, and of best available tube inspection techniques.

. Discussions were also held concerning the following issues:

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Minutes: Materials & Metallurgy and Severe Accidents November 5-6, 1996 l

I e limiting each degradation mechanism to 20% of the total probability of

failure criteria, and i

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projecting crack growth rate without leaving flawed tubes in service.

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! Regulatory Guide Leakage Monitoring and Risk Guidance: Mr. Joseph Donoghue,  ;

) NRR, presented the proposed objectives, recommended action levels, and l l suggested tr.chnical specification limits for primary coolant leakage j monitoring. He explained the following regulatory guide alternatives for 4

addressinc severe accident steam generator risk:

e I show a low frequency (10 evsnts per reactor year) for events leading to potential tube thermal failure, demonstrate low conditional tube failure probability (about 0.1), or

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+ implement a combination of plant and procedural modifications to reduce

[g analyzed risk below either of the previous two alternatives i Mr. Donoghue and the Subcommittee Members discussed the confusing presentation i in the regulatory guide of the leakage performance criteria, and the basis for j the 150 l j the 10', gallon eventsper perday (gpd)year reactor leakage limit. They also discussed the basis for criterion.

j Dose Assessment Guidance: Mr. Jack Hayes,.NRR, presented the radiological

! dose aspects of the rule and regulatory guide. He summarized the proposed- )

i optional flex program, which would establish dose criteria based upon the i probability of accidents rather than the probability of iodine spikes. Mr.

j Hayes explained the relationship of the proposed changes to existing j regulatory dose criteria and to the technical specifications. He presented

the flex program benefits, which include flexibility of operation,

independence from degradation mechanisms, and elimination of the need for i completing a dose assessment each cycle. Mr. Hayes and the Subcommittee i Members discussed the reason for not using total dose equivalent.

i j INDUSTRY PRESENTATION l i

Introduction:

Mr. Richard Pearson, Northern States Power, introduced the i topics of the industry's presentation and summarized the industry's 4

perspective on the proposed rule. He stated that the industry was in general j agreement with the performance-based steam generator degradation management

] concept and the regulatory guide performance criteria. Mr. Pearson noted, f however, that, the industry disagreed with the amount of detail and some items i in the regulatory guide, such as:

• assessing potential degradation mechanisms and resultant inspection requirements,

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!' Minutes: Materials & Metallurgy and Severe Accidents November 5-6, 1996 l

i e the supplemental performance demonstration,

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• the treatment of deterministic uncertainties, e

the 20% limit on probability of burst for one mechanism, and

) e using the operational leakage limit of 150 gpd as an enforcement trip

point.

! Mr. Pearson explained that NEI was developing an industry guidance document to

implement the proposed rule. He summarized the changes made to the NEI l industry document as a result of comments made by the NRC staff. He outlined

the content of the document and summarized the intent of each chapter. Mr.

j Pearson concluded that degradation specific management should be implemented I

in a timely manner, that the proposed rule has significant impacts on

} utilities, that there are many areas of agreement between the staff and i industry, and that the process will facilitate industry standardization.

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Technical Basis for Deterministic Structural Limits

Mr. Tom Pitterle,.

Westinghouse Electric Corporation, presented the differences between the

regulatory guide and the NEI industry document. He argued that the ASME code l elastic analysis rules could be interpreted to imply that a normal operational j structural margin limit of 2.6, instead of the factor of 3 proposed in the j proposed regulatory guide, is consistent with the ASME rules.

j Mr. Pitterle presented qualitative and probabilistic agreements for replacing i AP., differential pressure across steam generator tube at normal operating i

press, with AP d accidentcondiffo,ns.ifferentialpressureacrosssteamgeneratortubesduring He concluded that the operating structural limit presented in the regulatory guide should be 1.43 APace based on use of a burst correlation instead of the 3.0 AP, suggested in the regulatory guide, i

i Mr. Pitterle and the Subcommittee Members discussed the temperatures at which .

! tube burst pressure experiments were conducted and the correlation of j experimental results to tube at higher temperatures. They also discussed i graph 2-1 [ slide 1], " Distribution of Factor of Safety Against Burst," and the

inputs to the burst pressure calculation. The staff stated that it was j ' waiting for the industry report before reviewing the issue.

Conditional Probability of Tube Burst: Mr. David Steininger, EPRI,

! . recommended that the contribution of each form of degradation to the i conditional probability of tube burst criteria not be arbitrarily limited to 20% of the criteria. He presented the basis for the regulatory guide l conditional probability of burst criteria and introduced a risk relationship

equation [ slide 2] and assumed probabilities [ side 3] that supported his j recommendation. Mr. Steininger identified the different ways in which

{ defense-in-depth was used in the regulatory guide and concluded that the use j of a 20% limit on the conditional burst criteria for defense-in-depth purposes j was unnecessary.

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I . l Miriutes: Materials & Metallurgy and Severe Accidents l November 5-6, 1996 i

Mr. Steininger and the Subcommittee Members discussed the relationship between deterministic and probabilistic assessments, and the effect of summing the contributions of accidents other than main steam or feedwater line breaks to 4

the conditional probability of burst. The staff explained the necessity of addressing the contribution of more than one degradation mechanism to the

. probability of tube failure.

l Industry Implementation of the EPRI Steam Generator Inspection Guidelines: )

i Mr. John Smith, Rochester Gas and Electric Company, presented an overview of i

the development of the EPRI Steam Generator Inspection Guidelines. He explained that the industry would perter that the staff review the NEI Industry Document, and not the lower tier documents such as the EPRI Steam i

. Generator Inspection Guidelines. Mr. Smith explained how the utilities have implemented the EPRI inspection guidelines. He outlined the content of the  ;

i present revision of the guidelines including the preservice inspection

requirements and the important features of the industry performance i demonstration program. Mr. Smith summarized some of the changes contained in j the proposed revision to the guidelines. He concluded that the guidelines had i

improved inspections and noted that significant issues remained in the areas i j of the number of tube pulled required by the supplemental performance

, demonstration, validation of the inspection techniques, and the basis for a buffer zone.

I- Mr. Smith and the Subcommittee Members discussed the length of time required i to perform an inspection, the length of utility operating cycles, the

! applicability of the guidelines to replacement steam generators, and the

number of qualified inspection vendors.

SAFETY ANALYSIS PRESENTATION:

Introduction:

Mr. Donoghue, NRR, summarized the staff's approach to understanding the risk of steam generator tube failures. He outlined the staff's thought process, the example analysis, and the uncertainties identified in the analysis. In response to a question raised during the i June 3-4, 1996 Joint Subcommittee meeting, he presented the results of actual

, circumferencial tube failure events and concluded that there was no indication of a mechanically-induced tube failure resulting from the events.

Example Analysis: Mr. Donoghue explained that steam generator tube failures can be spontaneous or induced. He noted that spontaneous failures are not addressed by the rule. Induced failures can be caused by either pressure, thermal, or mechanical transients. Thermally induced failures can occur only during severe accidents where both the primary temperature and the pressure across the steam generator tubes are high. Mr. Donoghue explained that the

. staff performed an example analysis to understand the reactor coolant system conditions, which could lead to thermally induced tube failures, and the j reactor coolant pressure boundary response to those conditions.

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Materials & Metallurgy and Severe Accidents November 5-6, 1996 l i

Mr. Donoghue presented the accident progression event tree (APET) structure used in the example analysis. The staff calculated a frequency for all accident sequences that would result in high primary temperature and pressure i . coincident with a depressurized steam generator. The staff then constructed 3 an event tree to calculate a containment bypass frequency resulting from a a

steam generator tube failure. Mr. Donoghue explained the assumptions used to i construct the event tree, the derivation of the probabilities assigned to each i branch, and the result of the APET analyses. The results of the analyses are )

l summarized on slide 4. l l Mr. Donoghue described the following additional components of the example

. analysis:

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• conservative thermal-hydraulic analyses used to predict the relative times to failure of major reactor coolant pressure boundary components, i

j e the empirically developed flawed tube creep failure model, and l

• steam generator tube flaw distribution models developed independently by J NRR and RES.

1 l' Mr. Donoghue explained the assumptions use in the flawed tube failure j . probability calculation. He noted that the staff calculated the tube failure

probabilities of different flaw distributions for both pressure induced and ~

, temperature induced failures.

In response to issues raised during the June 3-4, 1996 Joint Subcommittee meeting concerning model uncertainties, Mr. Donoghue presented the results of several sensitivity studies. He described one study that evaluated the sensitivity of the flaw burst probability to various flaw distributions, creep model flaw multipliers, temperatures, and pressures. Mr. Donoghue identified the different cases used to evaluate the sensitivity of the APET. He i explained the set of sensitivity studies that evaluated the thermal-hydraulic model by varying the heat transfer correlations and the heat transfer coefficient between the counter-current flow of gases in the het leg. He .

described another sensitivity study that applied five percent limits to three parameters (mixing fraction, flow ratio, percentage of tubes carrying flow) in the steam generator lower plenum mixing model.

Mr. Donoghue noted that the key issues, which are associated with calculating i steam generator tube failures during severe accidents, were the representation of flaw distributions, the quantification of the event tree, the reactor  ;

coolant boundary weak points, the thermal-hydraulic results, and the tube i failure model.-

The staff and the Subcommittee Members discussed the following issues:

e cost benefit of the proposed rule,  !

. the use of split fractions for event tree probabilities, 3

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, Minutes: Materials & Metallurgy and Severe Accidents November 5-6, 1996 w sumed reliability of relief valves, e

use of the Larson-Miller parameter for predicting hot leg failure, transport and deposition of fission products in the secondary system, i

• effect of assumed source term on the calculated risk,

. flaw detection and characterization, defence-in-depth arguments, e

'; the relationship between tube temperature and tube failure probability, developing performance criteria based on offsite consequences, and the timing of primary coolant boundary component failures.

a Statement by Vendor Representative: Mr. Ray Schneider, ABB/CE, stated that the industry in response to severe accident concerns had developed strategies

and procedures for maintaining water in the steam generators and for depressurizing the primary coolant system. He noted that operator actions

) expected to mitigate severe accident consequences were not fully considered by

the staff. Mr. Robert Palla, NRR, explained that the staff did not consider

! operator actions due to the difficulty in assigning probabilities to the success of the actions and due to the unavailability of equipment needed to i- implement severe accident management ouiMine recomendations during the

uw.aed station blackout scenario.

j SUBCOMMITTEE DISCUSSION:

j Dr. Powers stated that the safety analysis for the proposed rule is extremely j complicated and that he was impressed with the work done by the staff. Dr.

Powers stated that he would like to better understand the amount of

! conservatism in the example analysis. Dr. Fontana stated the rule should be

reduced down to essentials, if at all possible, with application of as much j risk-based thinking as possible. Dr. Seale observed that the staff repeatedly

had to decide when to stop pursuing the different sets of analyses in order to i progress with the rulemaking.

F0LLOWUP ACTIONS:

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! The staff agreed to provide the Subcomittee Members copies of the following i docurents:

i Generic letter 95-05 and supporting information package,

• EPRI Document on Leakage Monitoring, and j

• Use of the Larson-Miller application for evaluation of hot leg failure l The Subcomittee Members suggested that the staff prepare a presentations for

the November 7, 1996 ACRS meeting, concerning the following:

an overview of the major blocks of the proposed rulemaking effort,

+ the issues and areas on which the staff desired Comittee guidance, j

• the statement that the regulatory guide is too prescriptive,

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., Minutes: Materials & Metallurgy and Severe Accidents November 5-6, 1996 hw risk is factored into the rule and regulatory guide, the flexibility. associated with the level of staff review proposed for i the different levels of industry documents, j e' the basis for the probability limits in the performance criteria, and

e the reasoning behind not looking at fission product transport and site specific qualitative health objectives as opposed to the conditional containment failure probability.

The Subcommittee Members suggested that the industry prepare a presentation for the November 7, 1996 ACRS meeting, which would include information on the industry's risk-based initial approach to the rule and the basis for the safety factor of 3 ed in the operational structural margin limit.

SU8COMITTEE RECOMMENDATIONS:

The Subcommittee recommended that the staff and NEI summarize the status of activities associated with the development of a proposed rule concerning steam generator integrity at the November 7, 1996 ACRS meeting.

BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE:

1. Memorandum date October 25, 1996, from Brian Sheron, NRR, to John Larkins, ACRS Executive Director,

Subject:

ACRS Review Of The Proposed Steam Generator Rule

2. Draft Steam Generator Rule (Revision 11.6) dated October 8, 1996 [ DRAFT PREDECISIONAL INFORMATION]

3. Draft Regulatory Guide X.XX, " Steam Generator Tube Integrity,"

dated October 8, 1996 [ DRAFT PREDECISIONAL INFORMATION]

4. Memorandum dated July 22, 1996, from Noel Dudley, ACRS Senior Staff Engineer, to ACRS Members,

Subject:

Certification Of The Minutes Of The ACRS Joint Subcommittee Meeting On Materials & Metallurgy And Severe Accidents, June 3-4, 1996

?RESENTATION SLIDES AND HANDOUTS PROVIDED DURING THE MEETING The presentation slides and handouts used during the meeting are on filed in the ACRS office and are attached to the printed transcript. Copies of the slides or handouts are available upon request.

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1 Minutes: Materials & Metallurgy and Severe Accidents November 5-6, 1996 NOTE: Additional details of this meeting can be obtained from a transcript of 4 this meeting available in the NRC Public Document Room, 2120 L Street, N.W., Washington, D.C. 20006, (202) 634-3274, or can be purchased from Neal R. Gross and Company Incorporated, Court Reporters and Transcribers, 1323 Rhode Island Avenue, N.W., Washington, D.C. 20005, (202) 234-4433.

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i Attachments: Slide 1 - Graph 2-1, " Distribution of Factor of Safety Against Burst.

Slide 2 - Risk Relation: hip Equation Slide 3 - Assumed risic values Slide 4 - Results Fron: Accident Progression Event Tree (APET) 4 4

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! :q ustry's Recommendation .

A nskrelationship shouldbe used

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u 1 4P[E}= P[R/E?+ P[F/E)1 P(D) < A E ( )

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Where, -

P(E), probability of the initiating avent, P(R/E), conditionalprobability of tube burst forunknown factors forthe initiating event, E, P(FE), conditionalprobability of tube burst for the degradation mechanism (j) being monitored, P(D), probability for failure to mitigate the etlects of the events andpreventcore damage, and A, value that wouldindicate an acceptable level of risk exists, frequency.

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The following values are used to show an acceptable level of dsk

• P(E) = 8.2E-4/yr.

- Value taken from SW94 NRC staffpresentation to ACRS for the sum of main steam line break and feedwaterline break frequencies p

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• P(R/E) = 0.1

- Value is assumed. It accounts for unanticipated causes of tube rupture during the faulted accident evident.

P(D) = 1E-3 }

- This is a typicalpublished value used by the NRC staff for evaluations of this type, e.g., NUREG-1477 T,,P(FIE) i.

= 0.05  ;

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- This value is the Rule's speci6 cation, but without the arbitrary reduction by 20%, and is allocated to all monitored degradation.

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RESULTS FROM APET ANALYSES o

Under base case assumptions, ~ l in 5 chance of containment bypass, given core damage with high primary pressure and dry SGs o

Majority of bypass events (90%) as";igned to large release category (RC-1 on APET)

Direct release to environment via open ADV or MSSV '

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No delineation in APET between single and multiple tube ruptures due to inability to assure integrity of adjacent tubes

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Ei.o Balance of events (10%) have lower releases due to partially depressurized RCS, intact (but leaky) MSIVs, and secondary side holdup / deposition (RC-2 on APET) j o

SBO with RCP seal LOCA is dominant contributor to thermally-induced SGTR frequency (70%), and total containment bypass frequency (50%)

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Conditional containment bypass probability found to range from 0.1 to 0.4 in preliminary sensitivity analyses (to be updated) minimum of 0.02 for optimal secondary side integrity maximum of 0.3 for 70K increase in temperature history ~

o Additional sensitivity analyses planned

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