NUREG-1570, Summary of ACRS Subcommittees on Matls & Metallurgy and Severe Accidents Meeting on 970304-05 in Rockville,Md Re Staff Draft NUREG-1570, Risk Assessment of Severe Accident Induced Steam Generator Tube Rupture

Summary of ACRS Subcommittees on Matls & Metallurgy and Severe Accidents Meeting on 970304-05 in Rockville,Md Re Staff Draft NUREG-1570, Risk Assessment of Severe Accident Induced Steam Generator Tube Rupture

ML20217H986
Person / Time
Issue date:	03/26/1997
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
RTR-NUREG-1570 ACRS-3051, NUDOCS 9708130366
Download: ML20217H986 (7)

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certif1ed: A3ril 3. 1997 issued: Marc 1 26. 1997 b21IFlJ = e, ADVISORY COMMITTEE ON REACTOR SAFEGUARDS JOINT SUBCOMMITTEE MEETING HlNUTES:

MATERIALS & METALLURGY AND SEVERE ACCIDENTS MARCH 4 5. 1997 ROCKVILLE. MARYLAND lhe ACRS Subcomittees on Materials & Metallurgy and on Severe Accidents held a joint meeting on March 4 5. 1997, at 11545 Rockville Pike. Rockville.

Maryland. in Room T 2 B3. The purpose of the meeting was to discuss with the NRC staff draft NUREG 1570. " Risk Assessment of Severe Accident Induced Steam Generator Tube Rupture " and the draft regulatory analysis. "SG Tube Integrity Rulemaking." The entire meeting was open to public attendance. Mr. Noel Dudley was the cognizant ACRS staff engineer for this meeting. The meeting was convened at 8:30 a.m. on March 4. and adjourned at 12:00 noon on March 5.

1997.

6TTENDEES:

(ESS R. Seale. Acting Chairman W. Shack. Member M. Fontana. Chairman 1. Catton Consultant T. Kress. Member N. Dudley. ACRS Staff D. Power. Member NRC STAFF B. Sheron. NRR S. Long. NRR J. Strosnider. NRR R. Palla. NRR J. Donoghue. NRR C. Tinkler RES T. Reed. NRR J. Hopenfeld RES INDUSTRY ,

C. Smith. Baitimore Gas and Electric Company R. Mullins. Southern Nuclear Operating Company There were no written comments received from the members of the public. One individual requested time to make an oral statement. An attendance list of members of the NRC staff and public is available in the ACRS office files.

Dr. Shack has a conflict of interest regarding the proposed steam generator rule and did not participate in the Committee deliberation on this issue.

INTRODUCTION: [.5 k 3/

Dr. Robert Seale. Acting Chairman of the Joint Subcommittee on Materials &

Metallurgy and Severe Accidents. convened the meeting at 8:30 a.m. He stated that the purpose of this meeting was to gather information concerning the

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technical basis and regulatory analysis associated with the proposed steam generator tube integrity rule. He noted that the staff was expected to address questions and concerns identified at previous Subcommittee meetings and in ACRS letters. Dr. Seale then called on Dr. Brian Sheron to begin the staff presentation. .g , g A tg s -b -

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o qtes: March 4 5, 1997 -2 M&M and SA Joint Subcommittee  !

STAFF PRESENTATION:  :

Introduction - Dr. Brian Sheron, NRR '

l Dr. Sheron provided the background, objectives, and motivation for the  !

aroposed steam generator tube integrity rule and associated regulatory guide.  !

le explained that the role of the proposed regulatory guide was to codify the  ;

acceptance criteria used in the current ad hoc staff reviews of license amendments for alternate tube repair criteria. He noted that the regulatory i guide describes a framework and criteria that licensee methodologies should  :

meet, but does not prescribe specific repair methods or limits.

Dr. Sheron stated that based on the results of the regulatory analysis, the I stoff-is reconsidering whether ruiemaking is the best vehicle for revising  ;

steam generator tube integrity requirements. He noted that some requirements in the proposed regulatory guide, such as, steam generator tube inspection programs, and structural and leakage assessments, can be justified as a compliance backfit. These requirements could be imposed on licensees through a generic letter.  :

The Subcommittee members and the staff discussed the consistency between the performance criteria in the proposed regulatory guide, the requirements in  :

generic letter 95 05. " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." and the guidance in draft regulatory guide OG 1061. "An Approach for Using Probabilistic Risk Assessment in Risk Infon,ied Decisions on Plant-Specific Changes to the Current Licensing Basis." They also discussed the following issues:

. whether the proposed regulatory guide allows for an increase in risk. {

e whether the end of cycle integrity verifications were being perforned by licensees or were included in technical specifications, e whether the staff would review licensee programs and methodologies ,

developed in accordance with the proposed r Julatory guide,

. -whether the presently available inspection techniques accurately characterize steam generator tube defects,

.-- whether crack growth rates can be predicted when tubes are repaired upon crack detection, and l

• the schedule for ACRS review of future staff activities.

Severe Accident Induced Steam Generatur Tube Ruoture Safety Assessment -

--Mr. Joseph Donoghue, NRR Mri Donoghue recap)ed the previous staff presentations to the Committee

.concerning the tecinical bases for the proposed rulemaking and explained the

', Minutes: March 4 5. 199'/ 3 M&M and SA Joint Subconnittee relationship between the draft NUREG-1570. " Risk Assessment of Severe Accident Induced Steam Generator Tube Rupture," snd the draft regulatory analysis. He presented the accident progression event tree (APET). which focused on events scenarios that occurred af ter a plant reached the conditions of high primary pressure and a dry secondary system. Mr. Donoghue compared the resulting core damage frequency of the APET analysis for the Surry facility to the individual plant examination results of a similar event for other pressurized water reactors. He then explained the assumptions used to evaluate the probability of each of the eight top level events in the APEl.

Mr. Donoghue presented the representative APET sequences, the release categories, and the changes made to the APET subsequent to issuing the preliminary draft of NUREG 1570. He described the possible design-specific influences and analysis uncertainties of the following:

. reactor coolant pressure boundary component failure probability.

. thermal hydraulic behavior, e relief valve failure characteristics, and e reactor coolant pump seal loss of coolant accident potential.

Mr. Donoghue identified the reactor coolant pressure boundary weak points that were analyzed by the staff. He provided detailed information on the results of thermal-hydraulic sensitivity analyses and tube failure probability calculations. He explained that the staff was conducting scoping estimates of steam generator tube damage resulting from degradation of known tube defects during severe accidents. Mr. Donoghue presented the analysis limitations resulting from the uncertainties associated with the steam generator tube flaw distributions, the flaw failure models, and the assumptions used in the accident progression event tree.

Mr. Donoghue stated that future revisions to NUREG-1570 will address the revised APET and associated sensitivity studies, the potential for loop seal clearing, the impact of design factors on thermal dynamic response, and tube failure model development that would account for circumferential cracks. He concluded his presentation by noting the following insights gained from the NUREG-1570 analyses:

e secondary system pressure integrity is as significant to the probability of core damage frequency as reactor coolant system pressure integrity, e plant- and design specific factors should be more fully considered, and e the small contribution from temperature-induced steam generator tube ruptures without reactor coolant pump seal failures indicates that current tube structural integrity criteria do not present undue risk.

The Subconnittee members and the staff discussed the uncertainties associated with determining steam generator tube flaw distributions, the consideration of scenarios that may result in reactor coolant system repressurization, and the number of licensee event reports related to tube failures. They also discussed the uncertainties related to the thermal hydraulic model including

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Minutes: March 4-5. 1997 4 M&M and SA Joint Subcomittee the heat transfer between the hot and cold fluid streams in the hot legs, the nodilization of tubes. the entrance effects on flow at the tube sheet and tube entrances. lack of silimitude in the scaling of experimental tube diameters, 1 and the potential for aerosols plugging tubes. The Subcommittee noted that i the pipe creep failure model for the hot leg was over simplified and did not '

consider the thickness of the pipe wall, nor the temperature gradients and j stresses around or along the pipe. '

The Subcomittee members and the staff discussed the bases for the assumed probability distributions used in the APET, the source of failure rate data, and the reason for selecting the new base case for the APET. They also discussed loop seal clearing as the dominate contributor to tube failure probability. the need to study circumferential cracks, the method the staff would use to review site s)ecific APET analyses submitted by licensees and the relationship between osservable crack lengths, and the crack lengths assumed in tube failure calculations.

The Subcommittee members and the staff discussed the need for severe accident induced tube rupture backfit requirements, the meaning of relaxation of requirements, and what constitutes a compliance backfit.

Steam Generator Rulemakina Draft Reaulatory Analysis -

Mr. Timothy Reed. NRR Mr. Timothy Reed. NRR. introduced the objective and structure of the draft regulatory analysis. He presented the problem statement, attributes of a proposed solution, and several viable regulatory solutions related to the steam generator integrity rule. He explained the requirements contained in 10 CfR 50.109. "Backfitting."

Mr. Reed explained that the regulatory analysis assumed that the risk from spontaneous and pressure induced steam generator tube ruptures would be unchanged by the rule, and that the risk from core damage induced steam generator tube ruptures would be reduced. Key cost related assumptions in the analysis included the cost of complying with the proposed rule, but did not included the cost of plant-specific probabilistic risk assessment calculations. The analysis included a net value ecuation that consisted of five factors for risk and two factors for costs. Fr. Reed provided a detailed explanation of each of the factors.

Mr. Reed noted that the staff is considering the preparation of a proposed generic letter which would include compliance requirements for odi fying technical specifications and voluntary requirements associated with implementing alternate repair criteria, The Subcommittee members and the staff discussed measurable and tolerable performance indicators. the derivation of the factor for aublic risk (VI), the discount values used in the net value calculations, and t1e net value that would support rulemaking. They also discussed managing risk, applying the concept of defence-in-depth, and developing quantitative criteria for imposing defense-in depth requirements. The example used was balancing risk by

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, Minutes: March 4 5. 1997 5 M&M and SA Joint Subcommittee allowing an increase in expected event frequency and an assumed decrease in event consequences resulting from operator actions, design changes, or application of.a revised sourre term.

The Subcommittee members and Dr. Sheron discussed staff plans for issuing the proposed generic letter. Concerns expressed by Subcommittee members included rewriting the regulatory guide to clearly identify compliance and voluntary requirements, and to focus on the severe accident induced part of the risk.

INDUSTRY COMMENTS: -

Mr. Rick Mullins. Southern Nuclear Operating Company Mr. Mullins, as a member of the NEl task force on the steam generator rule, requested that DG 1061, draft NUREG 1570, and the draft regulatory analysis be made available to the industry. He noted that there are several requirements in the draft regulatory guide, such as. NDE_ requirements for ins)ection techniques and some performance criteria. that exceed anything tlat the industry is presently doing. He disagreed with the staff )osition that these requirements could be jus +1fied as a compliance backfit. ir. Mullins stated that the o)erational assessments and conditional monitoring were being performed )y licensees and could not be separated as required and voluntary items.

PROFESSIONAL OlFFERING OPINION -

Dr. Joram Hopenfeld, RES Dr. Hopenfeld commented on the analyses presented in draft NUREG-1570. He stated that the results are invalid because leaks through degraded tubes are ignored. He compared the results of his analyses for containment bypass frequency and the degree of mixing in the steam generator inlet plenum with the results documented in draft NUREG-1570. He explained that his assumption of no mixing in the inlet plenum would lead to tubes failing before the surge line during severe accidents, which would result in a higher containment bypass frequency than that calculated by the staff.

Dr. Hopenfeld explained how the jet effect resulting from a pin hole tube leak could erode adjacent tubes and cause multiple tube leaks. He concluded that key assumptions in the NUREG report are incorrect and that the treatment of uncertainties does not cover the proper parameters.

The Subcommittee members and Dr. Hopenfeld discussed his calculational methods, assumptions, and test data used in modeling the jet effect: and the derivation of the flow out of the tube defects.

EBCOMMITTEE DISC _USS10NS:

Dr. Powers stated that he had a good understanding of what the staff understands about the risk of steam generator tube ru>tures and that the staff understanding is different than what he presumed in tie past.

( Minutes: March 4 5. 1997 6 M&M and SA Joint Subcommittee Dr. Kress. Powers, and Shack discussed the appropriate risk measure to use for evaluating the steam generator rule in accordance with guidance in DG 1061.

They discussed whether the measure of risk was bounded by latent or acute deaths, and whether the measure should include latent injuries.

Dr. Fontana stated that he did not understand the linkage between crack size and the vulnerability of tubes under various temperature and )ressure loads.

He stated that sensitivity and uncertainty analyses are not tle same.

Dr. Catton stated that the staff did not adequately perform sensitivity analyses on heat transfer coefficients used in the SCDAP/RELAP5 code calculations, in his opinion, the staff should justify the range of values used in the sensitivity analyses and deal with plus and minus changes to the different heat transfer coefficients in the same calculation.

Dr. Seale stated that the staff should allow the Subcommittee an opportunity to review any proposed generic letter or document associated with steam generator tube integrity prior to the public comment period. In particular, he indicated that the Subcommittee should review the staff determination of which items in the proposed regulatory guide are considered compliance and which are considered voluntary.

FOLLOWUP ACTIONS:

Dr. Sheron. NRR. committed to address the following items in separate letters or reports:

e concerns identified by the ACRS in past letters and memorandum.

e concerns identified by Dr. Hopenfeld in his differing professional opinion, and e resolution of generic safety issue 163. " Multiple Steam Generator Tube Leakage."

Based on the staff presentation the Subcommittee was satisfied that the follow items contained the November 20. 1996 ACRS letter, to the EDO had been adequately addressed by the staff:

. incomplete and sometimes perfunctory analyses required to provide an assessment of relative risk.

• reliance on core damage frequency as an indicator of risk Based on the staff presentation the Subcommittee was satisfied that the follow items contained in the January 31. 1997 memorandum, from John Larkins to Ashok Thadani had been adequately addressed by the staff:

• The staff should explain the basis for the 0.05 tube failure per year criteria -- in particular, how the value for team generator tube

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Hinutes: March 4 5.1997 7 M&M and SA Joint Subcomittee ruptures is an appropriate allocation of the total conditional containment failure probability; how the criteria for tube plugging are derived from inspection findings leak rates, or voltage criteria: or how plugging criteria are derived from the probability of tube failure.

. The regulatory guide does not explain how licensees should demonstrate that the syontaneous probability of steam generator tube failures will be below tie assumed failure criteria.

ElBCOMMITTEE RECOMMENDATIONS The Subcomittee requested that the staf f brief the full Comittee during the April 3 5. 1997 meeting, concerning the staff memorandum to the Comission regarding a new approach for addressino steam generator tube integrity. The Subcomittee requested that the staff 6e prepared to discuss its plans and to justify its new approach during the next Subcomittee meeting BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE:

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

Subject:

ACRS Review of the Proposed Steam Generator Rule (forwarded latest drafts of the proposed rule and regulatory guide]

2. Letter dated January 2. 1997, from James H. Taylor. Executive Director for 0)erations, to T.S. Kress. ACRS Chairman.

Subject:

Staff Response to ACRS Comments on Proposed Steam Generator Rule PRESENTATION SLIDES AND HANDOUTS PROVIDED DURING THE MEETING:

The presentation slides and handouts used during the meeting are on file in the ACRS office and are attached to the printed transcript. Copies of the slides or handouts are available upon request NOTE: Additional details of this meeting can be obtained from a transcript of 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. 0.C, 20005.

(202) 234-4433.

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