Meeting Minutes of the ACRS Materials & Metallurgy ... · Dr. William Shack, Vice Chairman of the ACRS Materials and Metallurgy Subcommittee, presiding, convened the meeting and stated

UNITED STATES NUCLEAR REGULATORY COMMISSION

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C. 20555-0001

April 22, 2003

MEMORANDUM TO: William Shack, Vice Chairman, Materials and Metallurgy Subcommittee

FROM: Ramin Assa, Senior Staff Engineer ~. /1~. Technical Support Staff / /

SUBJECT: CERTIFICATION OF THE MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON MATERIALS AND METALLURGY - FEBRUARY 5,2003, ROCKVILLE, MARYLAND

I hereby certify that, to the best of my knowledge and belief, the Minutes of the subject meeting

issued April 22, 2003, are an accurate record of the proceedings for that meeting.

Date

UNITED STATES NUCLEAR REGULATORY COMMISSION

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C. 20555-0001

April 23, 2003

MEMORANDUM TO: ACRS Members

FROM: Ramin Assa, Senior Staff Engine.er ~. 4 ~. Technical Support Staff /' /' (

SUBJECT: CERTIFIED MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON MATERIALS AND METALLURGY - FEBRUARY 5,2003, ROCKVILLE, MARYLAND

The minutes of the subject meeting, issued on April 22, 2003, have been certified as the

official record of the proceedings of that meeting. A copy of the certified minutes is attached.

Attachment: As stated

cc: J. Larkins S. Bahadur R. Savio H. Larson S. Duraiswamy S. Banerjee F. Moody V. Schrock ACRS Staff Engineers

· .

CERTIFIEID BY: W. Shack Certified on: April 23, 2003

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MINUTES OF ACRS SUBCOMMITTEE MEETING ON

MATERIALS AND METALLURGY FEBRUARY 5, 2003

ROCKVILLE, MD

INTRODUCTION

The ACRS Subcommittee on Materials and Metallurgy held a meeting on February 5, 2003, at 11545 Rockville Pike, Rockville, Maryland, in Room T-2B3. The purpose of the meeting was to hold discussions with representatives of the Office of Nuclear Regulatory Research (RES), relating to technical basis for revisions of the pressurized thermal shock (PTS) screening criteria in the PTS rule. Mr. Ramin Assa was the cognizant ACRS staff engineer for this meeting. The meeting was convened at 8:30 a.m. and adjourned at 4:50 p.m. on the same day.

PARTICIPANTS:

ACRS W. Shack, Vice Chairman V. Ransom M. Bonaca S. Rosen P. Ford G. Wallis T. Kress S. Banerjee, Consultant G. Leitch

NRC Staff D. Bessette J. Rosenthal E. Hackett N. Siu M. Kirk M. Mayfield

NRC Contractor A. Kolaczkowski, SAIC

There were no written comments or requests for time to make oral statements received from members of the public. A list of meeting attendees is available in the ACRS office files.

INTRODUCTION

Dr. William Shack, Vice Chairman of the ACRS Materials and Metallurgy Subcommittee, presiding, convened the meeting and stated that the purpose of the meeting was to review staff's draft NUREG report on the technical basis for revising the PTS rule (10 CFR 50.61.) Dr. Shack then called upon NRC staff to begin.

· .

Minutes: Materials and Metallurgy Subcommittee 2 February 5, 2003

NRC STAFF PRESENTATION

Introduction: Mr. Michael Mayfield, RES

Mr. Mayfield started his opening remarks by saying that the PTS Project has been a major undertaking for RES. He then introduced Mr. Siu and asked him to begin with the presentation.

OVERVIEW OF PTS RE-EVALUATION PROJECT - Messrs. Nathan Siu, Edward Hackett, and Mark Kirk, RES

Mr. Siu stated that this project has been supported by industry, specifically the Materials Reliability Program (MRP) of the Electric Power Research Institute (EPRI.)

Dr. Wallis noted that the draft NUREG report appeared to have been written by different people and was not well integrated. Mr. Siu acknowledged this point.

Dr. Ford questioned whether a thorough peer review was conducted as stated in the report's cover letter. Mr. Hackett stated that this activity was in progress and expected to be completed in 2003.

Drs. Ransom and Wallis pointed out that the NUREG does not provide a clear relationship between itself and referenced reports by University of Maryland and Oregon State University (OSU.) Mr. Bessette stated that the results of the OSU report were implicit in the NUREG.

Mr. Kirk noted that the objectives of the meeting were to review the draft NUREG and show a strong case to support rulemaking. Results of the plant-specific evaluation of two of the most embrittled plants in the fleet had shown that these plants had more margin against failure by PTS than previously believed.

Dr. Wallis stated that figure 1.1 in the report was very confusing. Mr. Kirk acknowledged and stated that the two sigma margins were misrepresented. Mr. Hackett added that there has been a fair amount of confusion over this issue over the years and RES' goal was to clarify this issue during the meeting. Mr. Wallis reiterated the need for peer review to identify and correct errors before issuance of the final report. Mr. Kirk acknowledged that the project was not over yet and needed additional reviews, including a response from Office of Nuclear Reactor Regulation (NRR).

Dr. Ford raised the issue of plants that were approaching RTPTS screening criteria and were interested in applying for license renewal. If the current 10 CFR 50.61 rule is not changed, these plants could not easily request a license extension. According to Mr. Hackett, Palisades Plant is the closest to and is projected to reach the screening criteria around 2011.

Mr. Rosen noted that the report only provides the technical basis for a change to the current PTS rule and asked about the criteria used for deciding whether to proceed with a rule change. Mr. Hackett responded that a petition for rulemaking from the industry could initiate this activity but the allocation of resources would the determining factor.


ANALYSIS APPROACH - Mr. Kirk, RES

Mr. Kirk presented a brief background of the PTS project. The licensee for Yankee Rowe power plant had predicted that the vessel embrittlement would reach the current PTS screening limit before the end of the plant's licensing life (EOL) and had attempted to follow the provisions of Regulatory Guide (RG) 1.154 to support operations at embrittlement levels greater than those implied by the screening criteria in 10 CFR 50.61. However, their efforts were not successful and the plant was permanently shut down in 1991. Following the difficulties with implementing RG 1.154, the Commission directed the staff to revise the RG and associated rule.

Since the original PTS rule was issued, improvements in Probabilistic Risk Assessment (PRA) analysis, thermal-hydraulics studies, and probabilistic fracture mechanics calculations, suggest that the current rule may be overly conservative. In the analysis supporting the development of the original rule, it was shown that a shift in mean value of the fracture toughness transition temperature to 210° F corresponds to yearly through-wall cracking frequency of 5X1 0-6

. A mean transition temperature of 201°F corresponds to a transition temperature of 2700 F computed following RG 1. 99, Rev. 2 because the RG 1.99, Rev. 2 temperature includes a margin term. The figure on page six of the handouts represented the distribution of plants that were close to the current screening criteria. Mr. Kirk stated that plants get closer to the RTPTS

limit by about one degree Fahrenheit per year of operation.

The staff selected Calvert Cliff, Oconee, Beaver Valley, and Palisades for plant-specific studies. These plants represent each of the major pressurized water reactor (PWR) manufacturers. Two of the plants were projected to be the closest to the current PTS screening criteria limit at EOL.

The staff's estimate of the through-wall cracking frequency starts with an events sequence analysis. This analysis defines both the combination of events (scenario) that can lead to a PTS challenge to the vessel and the frequency of such events. The thermal-hydraulic conditions associated with each scenario are determined using the RELAP Code. These analyses give the temporal variations of pressure, temperature, and heat transfer coefficient acting on the embrittled vessel. Probabilistic fracture mechanics analyses, based on linear elastic fracture mechanics techniques, were performed using the FAVOR Code. These analyses calculate the conditional probabilities with which through-wall cracks will occur. These conditional probabilities are multiplied by the sequence frequencies to obtain an estimate of the yearly through-wall cracking frequency. ..

The probabilistic fracture mechanics analysis treats the pressure, temperature, and heat transfer coefficient variation with time for each scenario deterministically. FAVOR takes as input the pressure, temperature, and heat transfer coefficient values versus time at the vessel surface, calculates the heat conduction in the vessel, and computes the resulting thermal stresses. The stresses are then used to compute the driving force for fracture. At the same time, FAVOR calculates a distribution of fracture toughness of material, which is dependent upon the temperature, the fluence, and embrittlement characteristics. Comparison of the applied driving force with the toughness distribution gives probability of fracture.


PRA ANALYSIS - Mr. Kolaczkowski, SAIC

Mr. Kolaczkowski provided an overview of the PRA modeling approach and the plant specific PRA models. He stated that the Oconee PRA model is the most complete one, relative to over cooling scenarios. The model identified one hundred eighty-one thousand two hundred fifty eight over-cooling sequences. The initiating event frequencies and equipment failure data in the model were based on industry generic data. The human reliability analysis (HRA) was initially performed by NRC contractors. The Beaver Valley model was the second one prepared by the staff and was simplified based on results from the Oconee analysis which showed that some scenarios were relatively unimportant from a through-wall crack frequency perspective. Palisades was the last model prepared by the staff. Because the Palisades IPE included PTS scenarios, the staff started with a pre-existing model and modified it. Unlike the other two cases, the licencee was the keeper of the model.

The results of the PRA showed that medium and large LOCAs are bigger contributors to PTS than previously taken into account in the 1980s when developing the original PTS rule. Recent analysis also showed that the thermal stress (or temperature) is more dominant than pressure.

During the meeting, there were considerable discussions between the Subcommittee members and the staff regarding operator action and assigning probability values to them. Mr. Kolaczkowski stated that for some over-cooling scenarios operator actions playa key role, either by mitigating or exacerbating the event. However, during a LOCA, which is the dominant event, operator actions have little impact. Thus, in PTS the uncertainties associated with operator actions have relatively little impact on the overall uncertainty in the vessel failure frequency.

THERMAL HYDRAULIC ANALYSIS - Mr. Bessettee, RES

Mr. Bessette stated that the staff used RELAP 51MOD 3.2.2 gamma Code to generate downcomer temperature, system pressure, and heat transfer coefficient at the inside of the vessel wall. These results were then used as input to FAVOR Code. Mr. Bessette presented a comparison between RELAP predicted temperatures and results of ROSA (Westinghouse) and MIST (Babcock &Wilcox) experiments. Members of the Subcommittee questioned the assessment of thermal hydraulic uncertainties, and their impact on the rates of change in the temperatures feeding into the FAVOR Code and asked the staff to present these results clearly and in more detail in the future.

PROBABILISTIC FRACTURE MECHANICS - Mr. Kirk, RES

The pressure, temperature, and heat transfer coefficient are input to an embrittlement and crack initiation model. Other inputs to the model include flaw distribution and their locations, orientation, material properties, composition, and fluence variations around the vessel. The model then calculates a yearly frequency of through-wall cracking. The flaw distribution data came from a variety of sources. According to Mr. Kirk, most of the big flaws (95 to 98 percent) are in the welds. Inspections have revealed that most of these flaws are fusion line flaws. This observation helps in the determination of the flaw orientation.


Generic Letter 92-01 required all licensees to report fluence level and identify limiting materials in terms of RTNDT' and characterize the embrittlement in terms of RTNDT. In addition, confirmatory experimental data were derived from tests at the Oak Ridge National Laboratory and other locations. The staff has recognized that RTNDT is not a precise representation of toughness changes under irradiation. However, even if better characterization of embrittlement were available for all materials of interest, there would still be aleatory uncertainty in the toughness. They developed a model describing both the epistemic and aleatory uncertainties in RTNDT and the aleatory nature of toughness, for both crack arrest and crack initiation. Dr. Wallis noted that the discussion of these uncertainties needed better clarification in the report.

Dr. Wallis asked about the effects of transients and flaw distributions in the stainless steel liner. Mr. Kirk responded that residual stress distribution due to the weld overlay and stresses caused by the differential thermal expansion of the stainless steel relative to the ferritic steel were incorporated in the analysis.

PLANT SPECIFIC STUDIES - Mr. Kirk, RES

Mr. Kirk stated that overall LOCAs are the dominant contributors to PTS failures in PWRs. There is at least three orders of magnitude uncertainty in through-wall cracking frequencies. Two thirds of the contribution come from the uncertainty in the LOCA frequencies and the remaining from uncertainties in the flaw distributions. The distributions are highly skewed and the mean and 95th percentiles are almost equal. Operator action does not playa significant role during most LOCAs because there is little an operator can do in response to it. However, for B&W plants operator action plays a more critical role in response to stuck open primary side valve scenarios. From a materials perceptive, the axial weld cracks and weld toughness or the plate properties dominate the RTNDT'

Dr. Ford questioned how could the results of this draft NUREG be applicable to all PWRs based on analysis of only three plants. Mr. Kirk responded that these plants were selected and ranked in terms of irradiation susceptibility and that because the challenge were dominated by LOCA events there is a high degree of consistency in operational challenge among plants.

ACCEPTANCE CRITERIA - Mr. Siu, RES

Mr. Siu described the reactor vessel failure frequency acceptance criteria development process. The strategy for developing the criteria was to be consistent with the original intent of the PTS rule by keeping the risk level low and keeping the relative contribution of PTS risk small compared to the risks associated with other sources. The staff believes that the reactor vessel failure frequency (RVFF) should be defined in terms of through-wall crack frequency rather than the frequency of crack initiation.

The key question was whether there is a margin between the occurrence of a through-wall crack and core damage and large early release associated with a PTS scenario. The staff urged that the challenge to the containment of PTS events is not exceptionally severe as compared to other accident scenarios. The important factor is the relatively low coolant temperature during a PTS events.


Mr. Bessette described several PTS transient scenarios. One scenario starts with a medium size LOCA, followed by a vessel failure in 1000 seconds. The FAVOR calculations and results of the pilot studies showed that the containment failure is unlikely and independent of a PTS event. Other scenarios also show that, overall, there is adequate margin between the occurrence of a PTS induced reactor vessel failure and large early release. For example, the reaction forces resulting from a vessel break are not worse than those analyzed for a cold leg break.

Mr. Siu concluded that the containment pressurization is likely to be less than a design basis LOCA and that choosing reactor vessel failure frequency criterion to be 10-6 would be consistent with the intent of the original PTS rule.

PTS SCREENING LIMIT - Mr. Kirk, RES

Mr. Kirk stated that the severity of PTS challenges is remarkably similar among the plants studied, and, the frequency of challenge is also fairly similar but with some greater plant dependencies. From a materials viewpoint, axial weld material and flaws dominate the throughwall cracking frequency and establish the relationship between the embrittlement metric and through-wall cracking frequency. Mr. Kirk described the RTNOT screening criteria graph which gives the relationship between the RTNOT and mean through-wall cracking frequency. The horizontal axis is the ASME RTNOT plus a shift due to irradiation calculated from the Eason formula. The vertical axis is derived from FAVOR calculations and incorporates all the complexities of uncertainties in material properties, thermal-hydraulics, and event frequency.

Using the graph, and taking the reactor vessel failure frequency criterion of 10-6, the resulting

screening limit RTNOT comes out to be 290°F. However, RTNOT is not the same as RTpTs .

Calculated by Regulatory Guide 1.99, Rev. 2, RTNOT is about 90°F less than RTpTs ' This suggests that a 80°F to 110°F increase of the current screening limit is possible.

CONCLUSIONS AND RECOMMENDATIONS

The staff stated that the purpose of this analysis was to show that a PTS event was unlikely and therefore the NRC could raise the criteria to allow the plants to run for a longer time. RES has forwarded the draft NUREG to the Office of Nuclear Reactor Regulation and believes that it can support revising the PTS rule. The staff highlighted that work on this analysis was still ongoing. The Subcommittee noted that the analysis and its conclusions apply to all PWRs. They also commented that the analysis and the draft report needed additional work and strongly recommended a peer review. The staff agreed that the draft report was not final and additional work was necessary. They also committed to present the information in plain language and clearer. The Subcommittee encouraged the staff to proceed with the rulemaking.

STAFF COMMITMENTS

1. The staff committed to perform a thermal hydraulic uncertainty analysis and evaluate the temperature distribution in the downcomer region.

2. The staff committed to perform additional FAVOR runs in terms of sensitivity studies.


4. The staff committed to revise the draft NUREG and perform a comprehensive peer review.

SUBCOMMITTEE DECISION

The Subcommittee decided prepare a letter regarding this matter and submit to the full Committee for consideration. The staff will brief the full Committee at the February 2003 ACRS meeting.

FOLLOW-UP ACTIONS

None.

PRESENTATION SLIDES AND HANDOUTS PROVIDED DURING THE MEETING

The presentation slides and handouts used during the meeting are available in the ACRS office files and as attachments to the transcript which will be made available in ADAMS.

BACKGROUND MATERIAL PROVIDED TO THE SUBCOMMITTEE

1. "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Criteria in the PTS Rule" (10 CFR 50.61), December 31,2002

2. Advisory Committee on Reactor Safeguards (ACRS) Letter to William Travers, "Reevaluation of the Technical Basis for the Pressurized Thermal Shock Rule", February 14, 2002.

3. ACRS Letter to William Travers, "Risk Metrics and Criteria for Reevaluation the Technical Basis Of the Pressurized Shock Rule", July 18, 2002.

4. William Travers letter to ACRS, "Risk Metrics and Criteria for Reevaluation the Technical Basis Of the Pressurized Shock Rule", September 3,2002.

**********************************************

NOTE: Additional details of this meeting can be obtained from a transcript of this meeting available in the NRC Public Document Room, One White Flint North, 11555 Rockville Pike, Rockville, MD, (301) 415-7000, downloading or view on the Internet at http://www.nrc.gov/reading-rm/doc-collections/acrs/ can be purchased from Neal R. Gross and Co., 1323 Rhode Island Avenue, NW, Washington, D.C. 20005, (202) 234-4433 (voice), (202) 387-7330 (fax), [email protected] (e-mail).

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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MATERIALS AND METALLURGY SUBCOMMITTEE MEETING

REEVALUATING THE TECHNICAL BASIS OF THE PRESSURIZED THERMAL SHOCK (PTS) RULE FEBRUARY 5, 2003, ROCKVILLE, MARYLAND

Contact: Richard Savio (301-415-7363, [email protected]) Ramin Assa (301-415-6885, [email protected])

-PROPOSED SCHEDULE

I. Opening Remarks W. Shack, ACRS 8:30-8:35 a.m.

II. PTS Re-evaluation Project Introduction M. Mayfield, RES 8:35-8:50 a.m. --,. {V00. f'.\."" Si",,

III. PTS Project Overview, Background M. Kirk 8:50-9:35 a.m.

Significance of RELAP differences wi D. Bessette 9:35-8:55 a.m. experiments (assessment results)

10:10-:1-2:00 a.m.

\~ .. Z~~ f'Wl

Plant Specific Results (Continued), A. Kolaczkowski, SAIC Applicability Beyond the study plants D. Whitehead, SNL Generalization and external events M. Kirk

R. Woods '2.0

V. Risk-Informed Reactor Vessel Failure N. Siu 2:10-3:1-Q-Frequency Acceptance Criteria D. Bessette Post PTS Vessel Failure considerations (including addressing comments of ACRS) Results of T-H analyses

VI. PTS RTNOT based screening limit M. Kirk 3:25-3:55

VII. Overall summary and conclusions M. Kirk, E. Hackett 3:55-4:&&-.36 )0 v,

VIII. Subcommittee discussion 4:ao~~. ..,,~

IX. Adjourn ?1-&p.m. NOTE:

• Presentation time should not exceed 50 percent of the total time allocated for specific item. The remaining 50 percent of the time is reserved for discussion.

• 25 copies of the presentation materials to be provided to the Subcommittee

---- -----=---~---.---.---.. --.--.-.----_saa:__ __

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS SUBCOMMITTEE MEE1-ING ON MATERIALS AND METALLURGY

FEBRUARY 5, 2003 Today's Date

NRC STAFF PLEASE SIGN IN BELOW

PLEASE PRINT

NAME NRC ORGANIZATION

(L~ /P I?-A A-/ EM ~

S:H 1tH M A-L1 k::-. I

J<oy LJo9~J {\E5 /nRAA/ PfAP .

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2810 Federal Register/Vol. 68, No. 13/Tuesday, January 21, 2003/Notices

"Auxiliary Feedwater System," to better reflect the four train auxiliary feed water [AFW) system design at STP. Specifically, the changes specify the same allowed outage time (AOT) for any one inoperable motor-driven pump, regardless of train. The amendments also extend the AOT for one inoperable motor-driven pump from 72 hours to 28 days. A sentence has also been added to Action d. stating that Limiting Condition for Operation (LCO) 3.0.3 and all other LCO actions requiring Mode changes are suspended until one of the four inoperable AFW pumps is restored to operable status. There is also an administrative change in the wording of the LCO to clarify that there are only four AFW pumps in each STP unit.

Date of issuance: December 31, 2002. Effective date: December 31, 2002. Amendment Nos.: Unit 1-146; Unit

2-134. Facility Operating License Nos. NPF

76 and NPF-80: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 22,2002 (67 FR 2930). The supplement provided additional information that clarified the application. did not expand the scope as originally noticed. and did not change the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 31, 2002.

No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: April 8, 2002.

Brief description of amendments: The amendments revise Technical Specification (TS) 3.4.16. "RCS IReactor Coolant System] Specific Activity," to lower the Limiting Condition For Operation and associated Surveillance Requirements for Dose Equivalent lodine-131 in the RCS from a specific activity of 1.0 ~Ci/gm to 0.45 ~Ci/gm.

Date of issuance: January 6, 2003. Effective date: As of the date of

issuance and shall be implemented within 60 days from the date of issuance.

Amendment Nos.: 102 and 102. Facility Operating License Nos. NPF

87 and NPF-89: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 11, 2002 (67 FR 40026). The Commission's related evaluation of

the amendments is contained in a Safety Evaluation dated January 6, 2003.


Virginia Electric and Power Company, Docket No. 50-338, North Anna Power Station, Unit 1, Louisa County, Virginia

Date of application for amendment: December 7, 2001, as supplemented by letters dated June 28 and July 25, 2002.

Brief description of amendment: This amendment permits a one-time extension of the current 10-year Title 10 of the Code of Federal Regulations Part 50, Appendix J, Option B, Type A test interval from April 3, 2003, to April 2, 2008.

Date of issuance: December 31, 2002. Effective date: As of the date of

issuance and shall be implemented within 30 days from the date of issuance.

Amendment No.: 234. Facility Operating License No. NPF-4:

Amendment changes the Technical Specifications.

Date of initial notice in Federal Register: April 30, 2002 (67 FR 21295). The supplemental letters dated June 28 and July 25, 2002, contained clarifying information only and did not change the proposed no significant hazards consideration determination or expand the scope of the initial application.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated December 31, 2002.


For the Nuclear Regulatory Commission. Dated at Rockville. Marvland, this 13th day

of January 2003.' . John A. Zwolinski, Director, Division ofLicensing Project Management, Office ofNuclear Reactor Regulation. [FR Doc. 03-1161 Filed 1-17-03; 8:45 am] BILLING CODE 7590-G1-P

NUCLEAR REGULATORY COMMISSION

Advisory Committee on Reactor Safeguards Subcommittee Meeting on Planning and Procedures; Notice of Meeting

that relate solely to internal personnel rules and practices of ACRS, and information the release of which would constitute a clearly unwarranted invasion of personal privacy.

The agenda for the subject meeting shall be as follows:

Wednesday, February 5,2003-1 p,m. until the conclusion of business

The Subcommittee will discuss proposed ACRS activities and related matters. The purpose of this meeting is to gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the full Committee.

Oral statements may be presented by members of the public with the concurrence of the Subcommittee Chairman; written statements will be accepted and made available to the Committee. Persons desiring to make oral statements should notify the Designated Federal Official named below five days prior to the meeting. if possible, so that appropriate arrangements can be made, Electronic recordings will be permitted only during those portions of the meeting that are open to the public.

Further information regarding topics to be discussed, the scheduling of sessions open to the public, whether the meeting has been canceled or rescheduled, and the Chairman's ruling on requests for the opportunity to present oral statements and the time allotted therefor can be obtained bv contacting the Designated Federal' Official, Mr. Sam Duraiswamv (telephone: 301/415-7364) between 7:30 a.m. and 4:15 p.m. (EST). Persons planning to attend this meeting are urged to contact the above named individual at least two working days prior to the meeting to be advised of any potential changes in the proposed agenda.

Dated: January 13, 2003, Sher Bahadur, Associate Director for Technical Support, A CRSIACNW. [FR Doc. 03-1221 Filed 1-17-03: 8:45 am] BILLING CODE 7590-G1-P

NUCLEAR REGULATORY

The ACRS Subcommittee on Planning \lOMMISSION and Procedures will hold a meeting on ..,,-AdVisory Committee on Reactor Febru~ry 5,.2003, Roo~ T-2B1, 11545 Safeguards Meeting of the ACRS Rockville PIke, RockVIlle, Maryland. Subcommittee on Materials and

The entire meeting will be open to Metallurgy' Notice of Meetingpublic attendance, with the exception of ' a portion that may be closed pursuant The ACRS Subcommittees on to 5 U.S.c. 552b(c) (2) and (6) to discuss Materials and Metallurgy will hold a organizational and personnel matters meeting on February 5, 2003, Room T

2811 Federal Register/Vol. 68, No. 13/Tuesday, January 21, 2003/Notices

2B3. 11545 Rockville Pike, Rockville, Maryland,

The entire meeting will be open to public attendance.

The agenda for the subject meeting shall be as follows:

Wednesday, February 5, 2003-8:30 a.m. until the conclusion of business

The Subcommittee will meet with representatives of the NRC staff and discuss the risk metric and criteria that can be used for reevaluating the technical basis of the pressurized thermal shock [PTS) rule and the NRC staffs pilot plant studies. The purpose of this meeting is to gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the full Committee.

Oral statements may be presented by members of the public with the concurrence of the Subcommittee Chairman; written statements will be accepted and made available to the Committee, Electronic recordings will be permitted only during those portions of the meeting that are open to the public, Persons desiring to make oral statements should notify the Designated Federal Official named below five days prior to the meeting, if possible, so that appropriate arrangements can be made.

During the initial portion of the meeting, the Subcommittee, along with anv of its consultants who mav be present, may exchange preliminary views regarding matters to be considered during the balance of the meeting.

The Subcommittee will then hear presentations by and hold discussions with representatives of the NRC staff, and other interested persons regarding this review,

Further information regarding topics to be discussed, whether the meeting has been canceled or rescheduled, and the Chairman's ruling on requests for the opportunity to present oral statements and the time allotted therefor can be obtained by contacting the Designated Federal Official, Dr, Richard P. Savio [telephone 301-415-7363) between 7:30 a.m. and 5 p.m. (EST). Persons planning to attend this meeting are urged to contact the above named individual at least two working days prior to the meeting to be advised of any potential changes in the proposed agenda.

Dated: January 14, 2003. Sher Bahadur, Associate Director for Technical Support, ACRSIACMV, IFR Doc. 03-1222 Filed 1-17-03: 8:45 am] BILLING CODE 759O-01-P

NUCLEAR REGULATORY COMMISSION

Peer Review Committee for Source Term Modeling; Notice of Meeting

The Peer Review Committee For Source Term Modeling will hold a closed meeting on January 28-29, 2003 at Sandia National Laboratories (SNL), Albuquerque, NM.

The entire meeting will be closed to public attendance to protect information classified as national security information pursuant to 5 U.S.c. 552b(c)(1) and as proprietary pursuant to 5 U.S.c. 552b(c)(4).

The agenda for the subject meeting shall be as follows: Wednesday, January 28 and Thursday,

January 29,2003-8:30 a.m. until the conclusion of business The Committee will review SNL

activities and aid SNL in development of guidance documents on source terms that will assist the NRC in evaluations of the impact of specific terrorist activities targeted at a range of spent fuel storage casks and radioactive material transport packages including those for spent fuel.

Further information contact: Dr. Andrew L. Bates (telephone 301-4151963) or Dr. Charles G. Interrante [telephone 301-415-3967) between 7:30 a.m. and 4:15 p.m, [EDT).

Dated: January 14. 2003. Andrew L. Bates, Advisory Committee Management Officer. [FR Doc. 03-1220 Filed 1-17-03; 8:45 am] BILLING CODE 7590-Q1-P

SECURITIES AND EXCHANGE COMMISSION

Sunshine Act Meeting

Notice is hereby given, pursuant to the provisions of the Government in the Sunshine Act, Pub. L. 94-409, that the Securities and Exchange Commission will hold the following meetings during the week of January 20, 2003. An Open Meeting will be held on Wednesday, January 22, 2003, at 10 a.m., in Room 1C30, the William O. Douglas Room, and a Closed Meeting will be held on Thursday, January 23,2003, at 10 a.m.

Commissioners, Counsel to the Commissioners, the Secretary to the Commission, and recording secretaries will attend the Closed Meeting. Certain staff members who have an interest in the matters may also be present.

The General Counsel of the Commission, or his designee, has certified that, in his opinion, one or more of the exem ptions set forth in 5

U.S.C. 552b[c)(3), (5), (7), (9)(B) and (10) and 17 CFR 200.402(a)(3), (5), (7), (9)(ii) and (10), permit consideration of the scheduled matters at the Closed Meeting.

The subject matter of the Open Meeting scheduled for Wednesday, January 22, 2003 will be:

1. The Commission will consider whether to adopt new rules 30a-3 and 30d-1 and amendments to rules 8b-15, 30a-1, 30a-2. 30b1-1, 30bl-3, and 30b2-1 under the Investment Company Act of 1940, amendments to rules 12b-25, 13a-15, and 15d-15 and Form 12b-25 under the Securities Exchange Act of 1934, amendments to Form N-SAR under the Exchange Act and the Investment Company Act, and new Form NCSR under the Exchange Act and Investment Company Act. These new rules and form. and rule and form amendments, would require registered management investment companies to file certified shareholder reports on new Form N-CSR with the Commission, and would designate these certified shareholder reports as reports tha t are required under sections 13(a) and 15(d) of the Exchange Act and Section 30 of the Investment Company Act. A registered management investment company's principal executive and financial officers would be required to certify the information contall1ed in its reports on Form N-GSR in the manner specified by Section 302 of the SarbanesOxley Act of 2002. The amendments would also remove the requirement that Form NSAR be certified by a registered investment company's principal executive and financial officers, and would provide that, for registered management investment companies, Form N-SAR would be filed under the Investment Company ,-\ct on1\' In addition, the amendments would implement Sections 406 and 407 of the Sarbanes-Oxlev Act by requiring a registered management investment company to provide disclosure on Form N-GSR or Form N-SAR, as applicable, regarding whether the investment company has adopted a code of ethics for the company's principal executive officer and senior financial officers, and whether the investment company has at least one "audit committee expert" serving on its audit committee, and if so, the name of the expert and whether the expert is independent of management.

2. The Commission will consider adopting rules to establish standards of professional conduct for attorneys who appear and practice before the Commission in any way in the representation of issuers. As proposed, the rules would require an attorney to report evidence of a material violation of securities laws, a material breach of fidUciary duty, or similar material violation by the issuer or by any officer, director, employee, or agent of the issuer to the issuer's chief legal officer or the chief executive officer of the company (or the equivalents); if they do not respond appropriately to the evidence, the rule would require the attorney to report the evidence to the issuer's audit committee, another committee of independent directors. or the full board of directors; if the directors do not respond appropriately, the rule would

ADVISORY COMMITIEE ON REACTOR SAFEGUARDS SUBCOMMITIEE MEETING ON MATERIALS AND METALLURGY

FEBRUARY 5, 2003 Today's Date

ATIENDEES PLEASE SIGN IN BELOW

PLEASE PRINT

AFFILIATION

E PiZ-.r W(?A~~~

A\I Cor-:l":..vL\l~G

ALsdr f-..\ kt, c.. Ae-.:c KG' u>..J1<J

DOtVt-/,e Wh~k~~J

YLl\\j- Hs)~ C~Ci±

1= SL

IS L

• l'

Reactor Vessel Failure Frequency (RVFF)

• RVFF criterion needed for two purposes:

• Su~port definition of RPV emtirittlement criterion

• Provide acceptance criterion for safety analysis

• Current metric and criterion established in RG 1.154:

RVFF=TWCF

RVFF* = 5 x 10-6 /ry

• Limited scope activi~ to revisit metriclcriterion in li~ht of recent risk-informearegulationinitiatives

Ci)O>.. °0 Q) C I wi>(1) ... :::JoC'" .... Q)o ...CUll. Q)

a:! >...2 'i: "coCUll.Q)

>

Vessel damage, age, or operational metric

VG 1

" .'

Task Activities

• Identification of options

• Scoping study of post-vessel failure accident progression

• Qualitative evaluation of technical issues

• Review of pilot plant calculations for T/H conditions

• Limited calculations

• Status reports and meetings

• SECY-02-0092 (5/10/02)

• ACRS (7/10/02), public meetings (10/17/02; 1/31/03)

• Chapter 5, draft NUREG (12/31/02)

• Focus on acceptabili~ => activities are largely independent of plant-specific studies .

VG 2

..

RVFF Acceptance Criteria Principles in Developing and Evaluating Options

• Consistency with intent of original rule

• Low risk level

• Low relative contribution

• Consistency with recent risk-informed initiatives

• Risk metrics

• Risk criteria

• Consideration of defense-in-depth

VG3

•

RVFF Acceptance Criteria Options (SECY-02-0092)

• Definition of RVFF

• RVFF =f(PTS-induced RPV through-wall crack)

• RVFF = f(PTS-induced crack initiation)

• RVFF acceptance limits

• .RVFF* = 5 x 10-6 / ry

• RVFF* = 1 x 10-s/ry

• RVFF* = 1 x 10-6/ry

VG4

..

Post-SECY Discussions

• Budgeting process: focus effort on assessingRVFF for pilot plants

• ACRS Letter (7/18/02; ML0220406120)

• RVFF should be based on considerations of LERF (and not CDF)

• Current LERF surrogate goal is not proper starting point

" ••.source terms used to develop the current goal do not reflect the airoxidation phenomena that would be a likely outcome of a PTS event."

• Options:

../' Develop acceptance criterion from prompt fatality safety goal

../' Use a frequency-based approach to develop RVFF* to provide assurance thatPTS-induced RPV failures are very unlikely

• ACRS' expectation: RVFF* will be substantially smaller than options proposed in SECY-02-0092 .

VG 5

;.

Definition of RVFF

It is appropriate to define RVFF as the frequency of through-wall cracks (TWCF)

• TWCF is a more direct indicator of risk than is the vessel cracking initiation frequency

• The current technology for predicting crack arrest is reasonably robust • Laboratory-scale experiments • Scaled-vessel experiments

VG6

Seoping Study - Key Questions

• Is a PTS-induced RPV failure likely to lead to melted fuel?

• Is a PTS-induced RPV failure likely to lead to a large, early release?

• Is the release spectrum (frequency-consequence) for PTS-induced large, early releases significantly worse than that associated with risk-significant, non-PTS-induced scenarios?

VG7

Scoping Study - Approach

• Refine SECY-02-0092 list of technical issues • Develop accident progression event tree (APET) to

support identification, representation and discussion of technical issues

• Evaluate current state of knowledge regarding technical issues

• Context for evaluations: \

• Focus on pilot plants; some consideration of plants addressed in generalization task

• Whether/how PTS changes accident progression

VG8

Accident Progression Event Tree (APET) u~

Q)Q)

c C u~guc Q) III 00co c

0 ;;: E ('-. E ('0. c U::O

.~ ·iii 0", .- Q) .- (J) c: 0 ~a.

c"o c 0> ~~

",,- -><c "0 III co >.-Sea ~,§&(J)C ;;: e 'Ee-g c "> t2"Ill c_I-~ e·~ f!x ~.~ .Qo 00 ~~ Ua.w uo uw J:(J) !DU- U!!l 8c%s: U-...J w8.s? (3t2

Arrested at. Circumferenllal Small .

Weld -0-10 in

Yes

-- -1- - - - - - -- -- --: - - - - -i-----f No

- Design Basis I

II i

... ...

I II ... RPV I

Very Larg~ i

Complete (»1000 In) i

i

2l c: Q) ~

tT Q)

(J)

1 2 3 4 5 6 7 8 9 10 11 12

13-24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40

187 188 189 190 191 192 193

III C ~ ~ .!!!,g ~~ m(l)~ 1!~~0:0 w 00:0 CI,l._:;s

.- Q) al·- Gl q.~~ = e-.!!? = ~ = ~8cr ~~cr ~~cr

Y

Y

Y

Y Y See Sequences 1-12

y'

Y Y

y'

Y Y

y'

Y y' Y

y'

Y y Y

y'

Y y' Y

Y y' Y

195 y'

"y" = yes 196 197 Y Y Y

"Y'" =yes, but reduced magnitude 198I I:,: : 194

199" "= no 200 Y Y YVG9

1

•

PTS 1Crackl Crack HOIE8low~tainrrtspray~ Fuel IECCSICavi~ i Eveti}jientaQ<tensi< Size Forc$olated!orkindocatio Runs~looded?!

1---1--------1---- ---I --------1- 1----2 3

Arrested at 4 !circumferenti1 Small 5

Weld 1-0-10 in'\ 6

8 9&1 1 1 1 1 I 7

Yes

110 11 12

"" 13-24I(-~'1:':'~"'11 !G- ---T -----r -----.r ------11--"_""_"_"_25 26

Beyond 27No Circumferen 28

tialWelds 29 30 31- Design 32 33 34

Basis

Large 35

~-100·1oo0 in' 36 37

Axial 38 » Design 39

1 Basis I ..• 40

InitiatesICircuC~~entrl

r k •••

I 1 ••• Arrested

Circum· 187ferential

I I- •• RPV 1 1 I 188 Very Large 189

Complete 1(»1000 in'') 190 Cavity I I I 191

192 193 194

RPV 1 1 I 195 196 197

Cavity I I I 198 199J I :~:':=:fs' but1ced magitude I I I ~ I I I 200

& «l E('o· ~~ > ._

~ ~ ~

(\jOO wua.

>~ ('0.

w 5l~Ol «l __ C»~ ~ «lOlQ....Ja:a.

c ~ 0 ~'ii('o. '1:g~Q) x._ o>Q ~

iii.!:::Q....J<l:a.

I I

Y

Y

Y

Y Y See Sequences 1-12

y'

Y Y

y'

Y Y

y'

Y y' Y

y'

Y Y Y

y'

Y y' Y

Y

y'

Y

y'

Y

Y

Y

Y Y Y

•

Potential Sources of Dependence Between Top Events

• Plant systems • RPV movement • Fragments • Fuel movement

VG 11

Plant Conditions at RPV Failure

• Power available, cooling systems running (injection mode)

• LOCA events: RCS cooling, depressurizing • MLOCA RPV failure at "'15-30 min (40 EFPY) • LLOCA RPV failure at "'5-10 min (40 EFPY)

• Stuck-open SRV events: RCS at SRV setpoint RPV failure at "'60-120 min (40 EFPY)

VG 12

Slowdown Potential After RPV Failure

,~ - ,::1212 F liquid 50lid G---EJ Oconee 109 (VFT =71405)

············l·············· .3e+11 . ,(j---E) Oconee 113 (VFT = 71405)

cr---t:" Oconee 122 (VFT =69605) ~ Oconee 124 (VFT = 43805)

......... ,[)-----b- Oconee 156 (VFT = 696052.5e+11

2e+11

1.5e+11 ~ . I... - - ....=)- - - - ....=...... - - '$ - - -C+- - - ~)- - - 8 -

, , , , , ,

r-···-·-·---1e+11 , ,

5e+10 ~ .

! , , Ia ;;.'~_--'--------.1._...L.-.--.J...._.L-3000 ' , --'---'-a 3000 6000 9000. 12000 1500 L- Til'!!l:!~s} ~ ~

VG 13


3e+11 ! , , , ,

... ... i-;212Fliquid solidG----EJ Palisades 40 (VFT = 300s) .... ~ Palisades 54 (VFT = xxxs) 2.5e+11

&---tI., Palisades 55 (VFT = 4620s) ~Palisades 58 (VFT =2700s)

, ,

.-.------------_ ..... 2e+11 ~

~

>- ..................... ,"':.:.: Q ; . .... . ':: :... . ~ --je> 1.5e+11 Q)

, . . ,~

1e+11 t.-..-..~.-..~.t\~.-'~.~.~.!.~-'.~.~-'.~r.-'.~ ..~.~~.~.~.-'.r.-'<3~.~.-'.~.j

:..--.,

5e+10 ;,--n ~:1\. n .=

L-' ~

O~l I I ! -------l! ~

-3000 0 3000 6000 9000 12000 1500C Time (~. . _

VG 14


8- - 8 212 F liquid solid , G---EJ Palisades 62 (VFT = 7205)

2.5e+11 ----------- ----------.- ----~- -------------: -- ----- -- - i--------------- ---I~ Palisades 63 (VFT = 15005) •--cr---t::, Palisades 64 (VFT = 19205) ~Palisades 65 (VFT =xxxxs)

2e+11 ~---------------------- ,-- --.------

----------------:Il--------------- -----,--1.5e+11 , , ,

---€)---'~' - -8- - - of)- - - 8 - - -& - - -€>- - - e- - -Go - - of)- - - 8 -

1e+11

"

5e+10 ~-" -------,,

o I I I I I

-3000 o 3000 6000 9000 12000 1500 ~ Time(s)___ I

VG 15


2.5e+11 !"-"-~-"""-~i

2.0e+11 ~---------------

I~ -I~) 212F liquid solid G---EJ BV 007 (VFT = 7205)

-------------,---- -------------I~BV 009 (VFT =3005) tr----f::. BV 056 (VFT =10505) ~ BV 097 (VFT =24905)

_.j._-----_ .. _------ •• -----;----_ ...• -------------_.:- ••• -._ •••• __ ••• - •• --.

1.5e+11 ~------------------

1.0e+11 I~--~:::I---

~.-----------------

. .. . .'- -~ ---c. --....-1- - ~ - - .-(~ - - ~ - -; - &Yts ts t ~ -6..... 6 .. ··-----------:--------- ...... -- .. -----,

5.0e+10 --.------------

O.Oe+OO L

-3000

~

o I ------'---'-_----'_~.L

3000 6000

IJ!!!E!Js) 9000 '. 12000

!

1500

. !

VG 16

RPV TH Failure Analysis

• ScopinR calculations performed using RELAPS/MOD3.3of RPY-Yailure for Calvert Cliffs

• Two transients analyzed - 4-inch surge line break - Stuck open gressurizer safety valves (2) that

reclose at 6 OOs • For each transient, two RPY failure modes analyzed

- 12 ft2 axial break (1 ft x 12 tt) - 3600 circumferential break

• For each break, three break opening times analyzed - 0.01 s - 0.1 s - is

• Results compared to Design Basis LBLOCA

VG 17

RPV TH Failure Analysis Circumferential Break Nodalization

Figure 1. Calvert Cliffs PTS Vessel Nading Diagram Circumferential Break

40 Six Azimuthal Region Annulus Inner Vessel

35 ~ta '''Hill LJ SIi. I", <lit hJAh. A""" l,.,,,. I 1!i6:l!1

,604t!i635 *:.. ~ i .I •30 250« • 18r"'rnT 2~ , 25n5T 1.. '" - .. - - _ _._ ..

... • , , I , 41142 2,165 • • 4 I •

• • • I •5 - - .. - - - : - - - - - - - -:- - . - - - - - -: - - - - - - - - ~ - - - - - - - - : - - - - - - - 25 .......... ·~·I·

,U •:~ : ·26 : ·21 :,·28 ~29 57101~c: 3,1ill8 Plln''\' ?al8 • • • I •0

+::0

1_ 6·_··~··!········~_·····_·~········~········!········1,_ ." • ;ir!l" • .a.~' ~.:

I • f·~ 20 7 - - - - - - - : - - - - - - - -:- - - - - - - - -~ - - - - - - - - ~ - - - - - - - -: - - - - - - - - .... , ;L:Q) I -I-.,, , • • I •

~ ••••••••••••••••p •••••••• " •••••••• " ••••••••••• _ •••••

o • •UJ 1_ • • • s"j ~ :Fl.• • 4 I • ~_....

15 ·_··_··r···-··_~·_··_····~·_--_··_~-_···_··,_··-·_·· li I a .I) 111~ 1_ fI--:; ::.'"(/l

8 I (; ":12 ... : $:t:~ Ulllll!l • _01.. . ...10 -M·~r--~j'

4,292

5 5.119

LllWI!I P","um 0

-5 dhl:f11JW

-2

VG 18

• • •

RPV TH Failure Analysis Axial Break Nodalization

Figure 2. Calvert Cliffs PTS Vessel Nading Diagram Axial Break

40 Six Azimuthal Region Annulus Inner Vessel

35 ~ta .RIIl., .'OJiH 1",1.'([" I"'JAii.. ...JA.Jii Lau,,, I Hi635

16635

~'" -.30 ,WI"2344

1~ 2~ : 25'1"""25T 1 ~ .•.•••••••. _.. _~.•••... _.p .•..••.. ~._ .. _- •••..•.. _.. ... , , , I , p~2._ • • 4 I •

• • • I •25 5 ~ - ;,; - .. : - .. - - - - -:- - .. - - . - -~ - ,- .. - .. -: - - - - - - . -; .,. - ... - - - ...........~

• ".25 ' -26 • ·21 '.28 '·29 51101·c 3.608 ~~r\, iUllll ~: : : : :

'.;:::l ·t·······~·········~········~····_···t········ 0

,? .~i~·..1.I·~J:Il111ll5 • • I I •~ 20 -.-&_._----~--_._----~----_.--~._-----_ .. _---_.- .... , :~:

Q) 1._ .I ~~--( , • _.. I '"

• • • I I iii ll111ll5 S" j e .:,~

• I • I •- .. r - -.- - -,. .. - - r • - •• - - •• , •• - •• - •• ll111ll5 ~-: & fj) 111391~ · .. .15 • PI ;/"'1::(/) -. - i - . - - -. - -:- - - --- . - -:- - - - - - -- -:- - - -- - - . - i . - - - - - - - l1--' iii 'r-''''U!985 10 · .. . 8 I ~ f{:!~

_ ... , .L_~ H- -··i--·----·~--··----i--------T-------·~·---·--· • ~'l'1_ I .

• • • I t • I · ..10 ·,········.· .. ·······{" .. ··· .. ·· .. r········,········~2' oM ·-'!il:l~r---:l1j"

4292

5 ~ 119

lo'Nef Pienum o

_5 I dR:<T.I04

-2

VG 19

I

Conditions at RPV Failure

Break Time Pressure

Downcomer Temperature

Specific Enthalpy

Transient (s) (psi) (F) (Btu/Ibm)

4-inch surge line break 2400 200 215

(saturated) 183

Stuck open SRV 8230 2400 355 327

lB loeA 0 2250 545 543

VG20

Internal Pressure Differentials

Core Barrel Downcomer Vessel ~p Core~P ~P Duration

Transient Break (psi) (psi) (psi)

4-inch surge Axial 10ms 150 60 150 12-30 ms line break Is 15 -10 25 ls

Cire 10ms 165 110 35 20-70 ms ls 45 30 15 Is

Stuck open SRV

Axial 10ms ls

1800 50

600 -10

1680 40

10-20 ms 130ms

Cire 10ms 2140 1460 50 10-20 ms ls 240 100 -15 60ms

LB LOCA N/A 10ms

ls 1010 -170

240 -70

1110 -500

VG 21

Containment Pressure

Containment Pressure Calvert Cliffs Vessel Breaks (ptscb02-1S) & LBLOCA (lblocb04-06)

~~ [' i , • , ' ii' I •• i , Ii. i , Iii iii i , , iii i , iii i • ii' i , i I' i , , i ' ii' J. i

1 ....... lbloea i-- p-llllCOlOOOOel 4 5 II -- jHililCO'lIlOO

-- jHililCOllllOO34 ~I --jHililCOlOOOO __ ptlllCOllllOO

33 11 -- I"I'lOOIOOOO

I!I ••••, p·I'lOOIOOOO ••••• jHililCOllllOO

31 ••• " jHililCO'lIlOO ..... I"flilOOlllOOO

30 ..... jHililCOllllOO ..... jHililCOlllOOO

32

29 __ jHililCOll1lOO

-- p<lllOOHll'..Q'ill 28 -- p.flQOOlllOOO .~ 27 Q;' 26 .. ? ~L-. Ci,cumf",,",lialllleak

:i 25III II 78

III 24 a:: 23

22 21

C<tCumfe,enlial bfeak11l,tU~.20 19 •• ," •••••.• ·"i<il's··iil·····•••.•••;";'j;iflieak" •••••,••••.

18 17 16 15 -5t I. i i • I I • i • I , 'i' Ii' ,I i. I • I i ' • ! I • " i I • i •• I • i i • 1

o 5 10 15 20 25 30 35 40 45 50 Time (5) t:I~)(ti.~

VG 22

•••

VG23

Key Sequences 'O£:: ~~

C C OLL. c c c: Ul 2l ;<

., EO'· '"E ". c: ~O c:.2 c"O c: C> ~C" Q)0 c c: ~c. ~~ 'ia~ :>"':'! c~ 19 ~:8.!!!'" ~~ c- o Gi~ 8~~ ie~ O.t! .20~i e'~ 0",

n.w 00 I Ow I :I:Ul I cou. I 8~ 8<%;;: u...J" 0 w8~ j~ l/l

Arrested al Circumferential.v~e1d

Va.

fBeyond Circumferential f'I~ 69

70 MedIum

•l 71

(-1G-100ln ) 72 73

No 74 Axial

Da.ign 75BaSIS 76Arrested ~r 77 78 79 80 81 82Large 83 84 85 86

~100-1000 in

BB 89 90

» Design 91 Basis 92

93

I I : i 87

Initiates I ' 94Circumferential 95

Crack

I I: ii 96

"y" =yes 100 101

"y'" = yes, but reduced magnitude RPV 102

" "= no 103 104

Cavil 105 106 107 108

RPV 109

Very Large LaViry: : jj~

Complete (» 1000 in/) - 114

RPV I I ~:~

i 117 118

. I ~~~

ArrestedCircum

~ 122 h:IHnJtlal :c:aV~iry~~~1~~~~iiiii 119

Complete IL..-----1 Rf'V ~ i iHICav'ly: - 126 127 128

8. ~~. i' ~.

<I:l q) wQ) CD

>-Of! ",l3:e'I:! ~ EJJ!! ~

~8~ ~~&£

V'

V

V'

V

V'

Y y'

y'

V Y

V

Y

yo

V

yo

V V'

yo

V Y

yo

Y

V

y'

V

Y

y'

V V'

V V'

V'

V V

V V

c i~~

'i~ CD

~1e ~~Q.

V

V

V

Y

Y

Y

V

V

Y

Y

V

V

V

V

V

•

Ql c:

PTS Crackl Crac~1 Ho~owao~tainniSprayi Fuel IECC~Cavi g r~ ~ Ql ~ ~~~ J Evetll ientad<tensi~r$ize IForc~olatedtorkiniPtatiq RunS1floode ; ~3: i~: ~;~

I -------11---------11--------11--------·1---------------"ArresieaafCiH:l.lmjereritial-\fJ~ld---

IBeyond Circutt~e~e~tial Wei

Y'~4%1 5%1 : -1 100 10') 71

• • • 72 Y Y Yes Axial I..&. . 73

D.eSlgn .. 11 74i J~P~_ W drlt--#===j ~~

Y'

YYArrested

T•~

--..~ ~~=j~~~77~ndent

Y' YNo -- 80 Y

IL-----tl:t==j 81 Large "'\----H------IIll 82 Y'

122-1000 in IL-----ft:t==j 8384 Y Y Y

~ r--::h-=-"' J----:-~t==:j: Y

~~H::j==:j 87

II

~I· 88 Y Y

IL-----ft~t=j 89 Y'+-H--~U 90 » Design

Basis ii-2 Y Y~laiiiiJiBti 91Initi~s

Circum! ential ,--~ ~ y' Cra k --Hif;::=tj 95I ~ Y Y' Y96.

Igp-Dependent I. 1~::1ij~~~9~9IY DepenCient

115

£~~~ltJt~-~-~~L~l!~:L_~_. l!~_11 -~~ I::; Y' Cavitv I ,"_4"_~_~_''" ~~~ Y Y' Y

120 121 Y Y' Y 122

RPV II If I 123 Y' 124

V·_~ __····~·I

Circum- 01 1ootI In01. 125 Y Y Y!erential :to ~:t0

Cavity II II 1126 127Complete 128 Y Y Y

VG 24

Observations

• Accident energetics are more benign than those of some other scenarios previously studied (e.g., HPME)

• Containment pressurization likely to be less than designbasis LOCA

• Blowdown forces on RPV and internals likely to be the same order of magnitude or bounded by DB LOCA

• Containment spray failure probability may decrease for PTS events (as compared with non-PTS risk-significantaccidents)* .

• Likelihood of fuel cooling dependent on reactor cavity design • Cavity flooding above top of fuel expected for some plants • For other plants, ECCS may not be sufficient to cool fuel

*For some plants, this may be dependent:on plant changes in response to GSI-191.

VG 25

· .

Scoping Study Conclusions

• The conditional probability of early fuel damage (given a PTS-induced RPV failure) appears to be • Extremely small for plants with cavities likely to be flooded • Non-negligible for other plants

• The conditional probability of early containment failure and a large, early release (given a PTS-induced RPV failure) appears to be very small for all plants

• Should a PTS-induced large, early release occur, such a release may involve a large-scale air-oxidation source term

VG 26

'"' .

Implications for RVFF*

• RVFF* =1 x 10-6/ ry is consistent with philosophy of original PTS rule, with ACRS guidance, and with Safety Goal Policy Statement • Assures a low level of risk associated with PTS events • Assures small relative contribution to acceptable risk • More limiting with respect to core damage than RG

1.174/0ption 3 criterion for CDF • Consistent or conservative with respect to QHOs

• Expectation: RPV embrittlement limits will be established in a risk-informed manner

VG 27

--Ulc: Ul ::J

o

-UCoU

~ to E E ::J

(J)

co N

~

Ul OJ

-C. -(J)

C. ::J ~ U co co

~ ) .

Technical Issues

• Definition of RVFF • Dominant plant damage states • Relative contribution of axial and circ welds • Crack propagation, hole size, hole location • Blowdown forces • Containment isolation • Missiles • ECCS status (injection, recirculation) • Containment spray status • Core status (intact, distorted, disrupted) • Fuel dispersal • Fuel coolability • RPV water level • Fuel environment (steam, air) • Early overpressure

VG30

••

.,. '~) lit

APET (Page 1 of 5) --. 0)

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•

PTS Re-Evaluation Project Briefing

Mark Kirk, Ed Hackett Probabilistic Fracture Mechanics (RES/DEr/MEB)

Sandia Nathan Siu, Roy Woods, Donnie~ Whitehead, Alar«olaczkowski III laboratories

National

Probabilistic Risk Assessment (RES/DRM/PRAB)

David Bessette Thermal Hydrauffcs (RES/DSARE/SMSAB)

ACRS Materials Subcommittee Meeting on PTS Re-Evaluation USNRC Headquarters. Rockville, MD • 5th February 2003

VGl

Topic '~entel"!l "Ille

OPE11 "9 P",m~r~~ 1='. Fcn1, AC>;'S 8:31)·8::36 ~ m

II. PTS R'H\'OIIUJUon Prole'~lancl Starr M. M.ytlElld, RES 8:3G·8:50 ~.m.

Imud l>:tions

III I'rs jtro,.C( 0 ......."..... 8acl"oIJnd M 1"J1~ . 8:50·9:55 am Signtlc3nce at RELe.p II fterente~ \"I: D. B:'SHItt e;o:perrrrent;.

IIREAK .:fi·l0:tO ••111.

IV. "lanllpulfj~ " ....II. (Ocon",,'1, Bego,,'er M. ~Jrk 10.10 a~l'

Val Eo/.I . l' all>acl""J 0 B:s~rtt~. 2: 11) IJT1 Th"rrn~I.H"l'd'<lijlr~ (.h~r.J~tenstic-s of or:mrm.rt R. \'VClOC~·,

If;sn~lenr£ urceolt3'IW,' reslIlt<; o VI111,tEt'r eacl. .e.. l<.oIJ tZ't'.'lsKI

LUNCH 1%;00-1:00 p.lII.

V RISk .Infor tn.IIIlHClOr '11"14'1 , all..,.. N 5iu. 1: 10.]: 10 Fr....n,.., AG••pUIK. CrltIri I D. B:sseltt Post PTS '.'~ss",1 ;:31u'l2 crt1~·der.3bor,s

I"MlJb ct T·H F1JI'(.sr-.s

BRI!AK 3:to~ze p.•.

VI PT:S RT.~ tH~·.d scr~",,,rng I'tnt M kirk 3: 26·3:'.i!5

VII. Oi>;!rall sLrnmary and con ~Iu;iorr. M. Mk E. H<o:Ketl 3:5G·"'I:55

VIII. S'JocorrrTlrt1ee dISClJS":;Cl 0 4:55-5: lOp m

AclJou rn 5: 10 p.rnIX

1

----

Broad Government and Industry Participation

_Dulcer61Energy. ~EAI

~~bFi!> Cn' .'f;t ,)I) {(':'

Sandia ~,;;!!!",,!!:~

~iiiiii . National /f..""""'1"1t''-, . Laboratories I ;RAMATOME"'.' ';." ~

lSL 0= SARTREX VG3

Meeting Objectives 1fTJR.[& ~?'1~

• Review Draft NUREG Tedlll~18lo919.r Rev19ltu(ae 1're9llU'tr.td

issued on 1231-02 110rntlal $JtM,1t(1"TS) SaenlJt;r: Criltria Ia 110< PTSR" (lIlCFR5Dii1.)from RES to NRR l~ _'_'_4'SO_~_ I

• Detail technical basis I~$~~-to support rulemaking

""-,~",,

CMM...... ,_,_·_·w ~.JI1T..: ....... 0.. ....• Discuss o ....going ..... ~... " _.. ................- .....

T_.~l ~.'&>

QIoI'I••activities --- -_. .. E-. ..........i:~':'l"'IIII .... _

• Address ACRS -~pJf1concerns

.I.'roI.",,~_11IIJ:• Request ACRS letter (l ..... J .._ ..~a......:-._

~;-~~~-

He

2

Conclusions

• These analyses provide a technical basis to recommend revision of the PTS rule • Two of the most embrittled plants in fleet have a

TWCF at or below 5x1(f at end of license extension (60 years)

• At the 10CFR50.61 Rllp' screening limits these plantshave a TWCF of 1xU'''lVS. RG 1.154 at 5xU')

• Analysis supports a revised screening limit of • 29C)'F on a weighted RIOT value

./ Axial welds & plates dominate

./ Cire welds and forgings minor contributors • This limit is 80F to 110'F higher than current

10CFRSO.61 limits on IQTs

VG 5

On-Going Activities

• RES activities • Calvert cliffs • Generalization to all plants • Sensitivity studies &. a more detailed examination of

current results • FavorV&.V • External peer review of project • Implications for operational limits (10CFR Appendix G)

• NRR activities • RES Draft NUREG sent to NRR on :1-31-02 • NRR comments due by 3U-03 • Decision to proceed with rulemaking?

vC.o

3

Briefing Overview

• 10CFRSO.61 (the PTS rule) • Background &. current implementation • Motivations for revision

• PTS re-evaluation project • Scope of analysis • Plant specific results

../ Analysis approach

../ Results

• Risk informed reactor vessel failure frequency acceptance criteria

• Conclusions ../ Rulemaking ../ Considerations regarding a new PTS screening limit

• On-going activities

VG 7

roun

4

10CFRSO.61 Back round & Current 1m lementation

lQCFR§5Q.61 If beltline materials are projectedSECY-82-465 Basis to exceed the RTNDTscreening limit

LONGITUDINAL CflACK EXTENSION NO ARREST

1C·2 c--~-,=.'CY="':,:::"'='"'=-rRE=.U=".,----r'----." at EOl, the licensee must either LEOf~O: Implement flux reduction and/or o "AA TOTAL perform vessel spedflc analysis too ST~ LIN! SAfAICS

t::. 5.0. TUBll RUPTURES justify continued operation.V 68LOCAW1WPS

¢' EXTENOED HI'I

• 10CFR 50.61: A mulN level structure

• Compare deterministicallycomputed RPV embrittlement (R1Ts)against screening criteria

• If necessary, employreasonably practicable flux

o reduction measures

....". .. Z2l5 2!0 27lIi _ ,. • If necessary perform plantspecific ana(ysis (RG 1.154)

MlA,H 8U"""(:1 I'lTflC,.rFJ

to justify continued operation

VG9

10CFRSO.61 (Motivations for Revision)

Yanl<ee Rowe

• In late 1980s the Yankee Rowe nuclear power plant was predicted to exceed the 10CFRSO.61 PTS screeningcriteria before EOL

• The Yankee Atomic Energy Compan~ followed the provisions of Regulatory Guide 1.15"4 in an attempt to build a case supporting operation to embrittlement levels beyond the screening criteria

• Yankee Rowe was permanently shutdown in Septemberof 1991

• 'rhe difficulties experienced with evaluation of the Yankee RG1.154 analysis led the Commission to direct the staff to revise the regulatory guide and associated rule

"" '0

5

10CFRSO.61 (Motivations for Revision)

• PRA • Use of latest PRA/HRA '<""" "7

data " / • More refined binning • Operator action

credited

• Acts of commission considered

• External events considered

• Medium and largebreak LOCAs considered

• TH • Many more TH

sequences modeled

• TH code improved

Technical Improvements made in the last20 years suggest conservatism of

the currel1t rule.

• PFM • Significant conservative biast

in toughness model removed • Spatial variation in fluence

recognized

• Most flaws now embedded rather than on the surface, also smaller

• Material region dependentembrittlement props.

• Non-conservatisms removed in arrest and embrittlement models removed

+ .t

State of art analysis methods adopted throughout VG 11

10CFRSO.61 Some plants '"close''' to the(Motivations for Revision) current screening criteria ~

licensee exemption requests ithout a systematJcprocess to ..

250

tJ) E 200 t-:J D. C) 150 Ec 0'- 100 <> ... c'l-CU LL e 50 o CJ

tJ) 0

2000 2010 2020 2030 2040

End of Current 40 Year License

6

Scope of Analysis

• All PWR manufacturers • 1 Westinghouse • 2CE • 1 saw

• 2 I?lants from original (19805) PTS study

• 2 plants veryclose to the current PTS screening criteria

• All potentialinitiating event sequences considered

Analysis Approach

7

Analysis Approach 2 main components Acceptance Criterion for • Plant TWC estimates TWC Frequency

Consistent with• Acce table TWC fre uen • 1986 Commission safety

~---P-LA-N-T-TW-C-E-S-TI-M-A-T-E-S----nll.~:~~~~~c::::~ement Uncertainties addressed and quantified as an • RG1.174

integral part of the analysis process

Details of PRA Event Sequence Analysis

8

Step 1: Collect Information

• Started with previous PTS PRA analyses • NUREG/C..3770 (Oconee) • WCAP-15156 ("Beaver Valley") • NUREG/CH183 (H. B. Robinson) • NUREG/C..4022 (Calvert Cliffs)

• Collected p-Iant s~eci'fic information for three plants analyzed (Oconee, Beaver Valley, and Palisades). Examples include: • Emergency and abnormal op-erating procedures, including

PTS relevant training material • Plant design information, • Existing PRA documentation, • Observed simulator exercises

• Periodic interactions with and feedback from licensees

VG 17

Step 2: Identify Scope & Features of PRA Model

• Initiators • LOCAs: small, medium, large • Transients: all types including support system

initiators • SGTR • SteamlineBreaks: small, large

• Types of accidents • Overcooling with lowering or otherwise controlled

RCS pressure • Overcooling with high RCS pressure • Overcooling withrepressurization • RCS faults, secondary faults, and combinations of

RCS a. secondary faults • At full power and at hot zero power

VG1'

9

Overview of Accident Scenario Modeling General Functional Event Tree for PTS

Initiator Primary Integrity secondary Pressure Secondary Feed Primary Flow/Press ok not PTS (1)

",o",kJ",c:"on",l"ro"lI",ed"---.,--.."...__minor PTS at most loverf.ed/pressurlzedl

"Oc;k -t0"'vc.:e"'''.::.ee'''d'- F1'n.:..o.:.."o'-'W'- possible significant PTS

1~lu:::n::;de::".:::ee=dJ:::lo::S:,1 core damage; not PTS

underfeed/lost .go to PMmalV Integrity failed (Feed & Bleed) (2) ok

r"0:::kJ"'C:"on:::l::ro~lI::ed::...=~--mjnorPTS at most lonrf..d/preasuriudl

not isolated/overfeed Ino flow possible significant PTS -depressurizing II."'U"'n"'de"'''"ee'''d''''"IO"S'-I core damage; not PTS

underfeed/lost go to PMmalV Integrity failed (Feed & Bleed) (3)

.... ...,I5Oe note (4)

(1) not considered a PTS concern regardless of primary flow/pressure (2) loss of feed to both SGs; procedures call for Feed & Bleed which is equivalent to entering tree at

Primary Integrity "failed" (3) like (2) above except secondary depressurization has further lowered ReS temp (4) logic is identical to rest of tree above except choices also exist for Primary FlowIPressure even for Secondary Pressure and Feed "ok" state and PTS effects are generally potentially greater for all scenarios

VG 19

Step 2: Identify Scope & Features of PRA Model (Continued)

• Operator Actions • Successes • Errors of omission • Acts of commission (proceduNiriven)

"e ,n

10

Classes of Human Failures Primary Integrity Secondary Secondary Feed Primary

Control Pressure Control Control PressurelFlow Control

• Operator fails to • Operator fails to • Operator fails to • Operator does not isolate an isolable isolate a stoplthrottle or properly LOCA in a limely depressurization properly align feed in throttlelterrninate manner (e.g., close a condition in a timely a timely manner injection to control block valve to a manner (overcooling RCS pressure stuck-open PORV) • Operator isolates enhanced or • Operator trips reactor

• Operator induces a when not needed continues) coolant pumps LOCA (e.g., opens a (may create a new • Operator feeds (RCPs) when not PORV) that depressurization wrong (affected) SG suppose to and/or induces/enhances a challenge, lose heat (overcooling fails to restore them cooldown sink...) continues) when desirable

• Operator isolates • Operator • Operator does not wrong path/SG stops/throttles feed provide sufficient (depressurization when inappropriate injection or fails to continues) (causes underfeed. trip RCPs

• Operator creates an may have to go to appropriately excess steam feed and bleed & (modeled as leading demand such as possibie overcooling to core damage opening turbine that way) rather than a PTS bypass/atmospheric concem) dump valves

VG 21

Step 3: Construct PRA Model

• Oconee and Beaver Valley • Event tree- small fault tree models used for both

power and hot zero power conditions

• Palisades • Event tree- fault tree, where fault trees

incorporated more component detail • Power and hot zero power combined in same model

He"

11

Oconee PRA Model Development

• First model to be constructed by NRC contractors • HRA initially performed by NRC contractors with

review by licensee • Initiating event frequencies and equipment failure

data based on industry generic data • No preliminary TH or PFM information available

during initial model construction • Hence, modeled "all" over cooling scenarios

VG 23

Beaver Valley PRA Model Development

• Model developed by NRC contractors using lessons learned from Oconee analysis • HRA initially performed by NRC contractors with

review by licensee • Initiating event frequencies and equipment failure

data based on industry generic data

• Utilized results from preliminary TH and PFM information

• Therefore, PRA model could be simplified

12

Beaver Valley PRA Model Simplifications

• Sequences involving: • Certain combinations of stuclDpen pressurizer PORVsor

SRVs were not modeled • Certain combinations of secondary valve and simultaneous

pressurizerPORV/SRV stucltopen events were not modeled

• Only secondary valve (single or multiple) stuclpen events were not modeled

• Only a single SG overfeed from AFW were not modeled • Secondary depressurization downstream of ttMSIVs

were not explicitly modeled • Steam generator tube ruptures were not modeled including

even ttiose involving lade of proper feed control and even with RCPsshutdown (possibly inducing RCS loopstagnation)

VG 25

Beaver Valley PRA Model Simplifications (Continued)

• Other sequences were screened from modeling on a case-by-case basis if the sequence frequency could be conservatively estimated at lower than "'1E-8/yr • Justification: • When coupled with the highesCPFs being

calculated for any type of sequence (in the:!: range), this woufd field a thNlall crack frequency of <E-11/yr range (thus would clearly not be important to the overall PTS results since some other sequences were known to involve thpwall crack frequencies in the f!IJ/yr range for reasonable EFPYs).

13

Palisades PRA Model Development

• Started with licensee's prexisting Palisades PRA model

• Modified by licensee to include NRC contractor input

• Collaborative HRA effort • Utilized initiating event frequencies and

equipment failure data contained in licensee's model

VG 27

Step 4: Quantify and Bin Modeled Sequences

• Individual accident sequences quantified • Combined "like" sequences into preliminary TH

bins • Developed new TH bins as necessary (an iterative

process) • Quantified pointestimate frequencies for all TH

bins

14

Step 5: Revise PRA Models and Quantify

• Models and preliminary results reviewed by • Licensees • Internal project staff

• Purpose of reviews was to determine: • Whether inaccuracies existed in the modelsr and whether

additional potential PTS sequences needed to be modeled, • Whether additionallH bins should be created, • Which human actions should be reexamined to produce

even more realistic (i.e., less conservative) human error probabilities fiEPs), and

• What combination of the above that could be accomplishedwithin the constraints of the project.

• Models were modified andequantifiedonthe basis of these reviews

VG 29

Step 6: Perform Uncertainty Analysis

• Each scenario (TH bin) is the interaction of what is treated as random events: • Initiating event • Series of mitigating eqUipment successes/failures

• Operator actions

• SO, the occurrence of each scenario is random FrequenClcenario= FrequenC1lnitEventX Probabilit}l;q.iPResponse x ProbabilitllpActiOnl

eillch withepistemic uncertainties described by a distribution

• The various scenarios. their frequencies characterize the aleatory uncertainties associated with the occurrence of a PT5 cliallenge

• Latin hyperc;;ube sampling techniq",es are used to propa«lin:~ th~pistef1]lc uncertainties to. generate a probaDlhty distribution for each scenario frequency

15

Step 7: Finalize Results

• Selected aleatory uncertainties were dealt with quantitatively • Size of the LOCA within a LOCA category plus other

factors (e.g.,initial injection water temperature), • Size of the opening associated with a single or

multiple stuck open SRV(s), • Time at which a stuck open SR\leclose!t and • Time at which operators take or fail to take action.

VG 31

General Form of the Results

set of 'F H Curves .,--_---.:I=-...:=.l1l:u:erta.i.n~ on each quantile estimate for Bin (Scenario)

I I I I

! PFM ~&PRA ..., Integration

Tasks o ,.. 2.2e-4 2.50-4 3,2&-4 1e-5

Time 1...~~I~!~~~~~~~~..p\ ......

Bin (Scenario) Frequency

(per year)

Sampling performed to quantify the epistemlc uncertainty in the bin frequency

• Histogram: 19 Quantile Levels (0.5%99.5%) plus maximum sampled value

• 95% confidence interval on each Quantile Vallie Lower llr. Upper Bounds

16

Thermal Hydraulic Analysis Approach

• Purpose of thermal hydraulic analysis: • Generatedowncomertemperature, system pressure and heat

transfer coefficient at the inside of the vessel wall for inji:atFAVOR.

• Code used for all analysis: • RELAPS/MOD3.2.2 gamma released in June 1999

• Applied previously developed models as the starting point: • Oconee- model dates from originallPTS study • Palisades- developed from model provided b'iemens Power

Corporation • Beaver Valley- W substantially revised H.B. Robinson IPTS model

to reflect Beaver Valley • Two-dimensionaldowncomermodel added and models revised to

reflect current plantietpointsand operating procedures

VG 33

Assessment of RELAPS for PTS Applications

• Assessment presented at the 12/11/02 Thermal Hydraulic Subcommittee meeting based on: • developmental assessment casesMarvikel) MIT

Pressurizer SemiscaleNatural Circulation, UPTF • integral test data: MIST, LOFT, ROSat, ROSAAP600

• Review and update assessment results from Subcommittee Meeting • Focus on Tests MIS1UOOB2",ROSAAP600 Test APCL-03,

AP-CL-09, and ROSAIV Test ::»&CL-18

• Show that: • RELAP5 provides good agreement fcdowncomer

temperature and system pressure • Effect of differences between code and experiment on

conditional probability of vessel failure

vr, 34

17

MIST Overview

• MIST (Multiloop Integral System Test) • Full height fulGressure, integral ~stem

experimental facili~ (power scaling factor is 817,volume scaling factor IS 620)

• BlkW lowered-loop design with two hot legs and fourcold legs.

• Major plant components modeled in MIST • Boundary systems provided simulation of the HPll

emergency feedwater, vents, controlled leaks, anasteam generator tube ruptures.

• Transient aSlessed is Test 4100B2 which is a 4.4 inch (100 cm) cold leg break

VG 35

MIST Results (4.4-inch cold leg break) 600 5S9 3000 207

500 533 :?500 17.2

400 478_ 2000 136 ~~ ~ .~ l;; !

~~ ~:. 300 4221 I!! 1500 103

,-J ill ·ii i 1 "-

1 366....200 1000 69

100 311 600 34

0 0 1000 2000 3000 400Il J5 o0 1000 2000 300Il 400Il soo8 0 Tiro<! (6) lim.. (6)

- MIST TB&!4 I 0002. 100 an' cold leg break(RELAP5) - MIST TB&!410002. 100 em' cold leg break (Experimental Datai

18

ROSA-AP600 Overview

• ROSA Facility • 1/30 Yolumescaled, full height, fu~lJressure

representation of a Westinghouse AP600 passive safety PWR

• Major plant components modeled in ROSA • Transients assessed:

• i-inch diameter break on bottom of cold leg (CII3) • AP-CL-09 - same as APCL-03 except with multiple

failures

VG 37

ROSA AP-CL-03 Results 600 5iIl9 3000 107

600 533 2600 172

400 473_ 2000 138 ~~ ~ .i :. :!?, i!

:> ! ~ :. 300 4221! I!! 1500 103!! 1 8. m i!

,.:i •e "-~

"-~

366+zoo 1000 6.9

100 311 500 34

0 0 2000 4000 6000 J5 2000 4000 6000 8008 0

lime ($1 lime (01

- ROSA-AP600. AP--CL-03. 1" cold leg break, 1ADS... 'ISIVB 'ail. to open (RELAP5) - ROSA.AP600, AP.CL.03. '" cold leg break. 1 ADS~ va)velail. to cpan (Experlmental Data)

19

ROSA AP-CL-09 Results 600 589 3000 207

500 S33 2500 17.2

400 478_, 2QOO 13,6 ~~ ?!- i l;!', !! ! ~ :il 300 =i!" l!! 1500 103!!!

ia i;\:~ 8E ~ ~ a. a.,-'. 200 366>-'" 1000 6,9

100 311 500 34

o0 2QOO 4000 6000 8000 1OOO~ o0 2QOO 4000 6000 8000 '00£0

Time (S"l Tif11(t (&}

- ROSA-AP600, AP-CL-09. I" cold log break. mtJ!I1plo sys_ loilures (RELAP51 _. Mf:;A",APfino. AP..CI ..nQ. ,~ N'\Id l~ lYP.<:Ik l'TUrttlpjA q;ystHn tAihJrAl:" ~F"~Flnt.Ri DRtl=!)

VG39

Effect of Differences Between RELAPS and Experiment

Case Mean Pressure Std Dev Error Mean DC Temp StdDev DC Error (MPa) (MPa) Error (K) Temp Error (K)

MIST 4100B2 -02 03 -4 11

ROSA -OA 0,7 -I 16 AP-CL-03

ROSA -0.1 02 0 9 AP-CL-09 ROSA-IV 02 02 -I 8 SB-CL-18

Mean values and standard deviations of the pressure and average downcomer fluid temperature error defined as:

Pressure Error ~ PRELAP - POATA Average Temperature Error~ TAVERELAP - TAVEoATA

An examination of the effect of these differences on conditional probabilities of vessel failure as calculated ~y

FAVOR is underway

20

PFM in the Overall Process

Conditional probability of yesse Yearly

failure CPF . frequency of thru-~all

•

VG41

PFM

PFM ..... .....-.

21

Probabilistic Fracture Mechanics SUlTlmary • Toughness

• Referenced to toughnesldata a. physical understanding

,/ Significant conservative bias in ultirradiated index temperature removed

,/ Non-conservatism in arrest model removed

,/ Aleatory nature of toughness uncertainty quantified

• Embrittlement • Referenced to toughness

data&: physical understanding

,/ Correlation with better empirical/physical basis

,/ Slight biases in in CVN based shift estimates removed

VG43

• Fluence • Spatial variation in

fluence recognized, significant conservatism associated with max fluence assumptionremoved

• Flaws • Based on significantly

more data than before • Most flaws now

embedded rather than surface flaws

• More flaws than before

100RTNDT Bias Correction ...---------------------1 ~

• Quantify how far 0"_-----..... ! 100

RTNDI' is from an accurate representation of real toughness data • Using a consistent

representation of that data

• To best represents "true" fracture toughnesstransition data

• Adjustment based on CDF ofaRT=RT NDT(u) - To ~

• aRT accounts forall known epistemic uncertainties

00 Uncertlrinty InRTf'lor relatlvelo

o +--~~~~d'~"~ ·200 ·100 • 100 200 300

100% ..----------~

._~ 80% :c.. e 60% I ___JI ;1,;111 :11 , ~

~ ~ 40%

.~ 20%U

0% +---=-..=--~_7A==+----J

-50 200

22

- -

® MoreflawsFlaw Distribution © Flaws smaller © Most flaw. ~uried

1.E+03- ...----------1 © Weld flaws on fusion line 0 0 1.E+02u. -- Base metal flaws 0

-- Surface/Clad Flaws :c 1.E+01C'll

~ -- Marshall(J A ... .r:::: 1.E+00 CI)a. a. III CI) 1.E-01~o C'll C) - cLL .- 1.E-02 -o >C'll ... .r:::: 1.E-03 CI) .c E 1.E-04

Z ~

1.E-05

0 5 10 15 20 25 30

VG45 a = Flaw De th Percent Wall

... . 5..,---------------,

at mid core (h=72 .. above bottom Of ..n,.'

N~4 ' /\ \~. !~ ~; ,:' ;

~o 3", \ \ •. .;," ~ ~ ~ i. f ~ ~ ~ \ ... ~ 2 ~ ." I ::'; :' l ~

5i 11\; ~\ /.1' \ I A \. ~1 V~ ~VJ~ VV ,g::I '"13 .. above bottom Of active core ~ 0 -"-..-..--"-'--------4}/".......~-'I ~ l' above top of active ?ore

~, below bottom of active core

o 45 00 135 180 225 270 315 361

azimuthal location (degrees)

Fluence

Previous (circal,aOs) PTS analyses .ssumet .... . 5.,-----------------,

at core flats fO 90 180 270 degrees)

\/..•- \N~4

E ~.. "'.

~ ~ 3 : a122.5, 67.5,112.5,157.5, Cl .: 202.5,247.5,292.5, 337.a. .. ~ //egrees ' .. 2

..u c::I =1 .<~::::::::::~::::::::~:~:~~:~::::::::~.:~~::~:.,..\, ,C g f/'"'" /''}-,.' ", '. ::I ~I~"" ..•.:~~; ~ 0

at 45,135, 225, 315 degrees

o 1224 36 48 60 72 84 96108120132144156168

axial location along core (inches)

23

.Previous analyses Oconee 1 Material Map assumed most embrittlement

00 90 0 1800

sensitive material to exis~_t I I-- __ I

evervwhere 12" 1" WeldelR 1135 .Weld.. 1073 Weld.. 1073

19.2" >-- Plate C21S7 6=1120 --.. 6=2920

--..

RTNDT ~"'~"~ 83"F @ EOL rACTIVE 9= 1360

>Plate C326573.2"CORE C3278Weld AX 1493

144" RTNDT 6~~ 151DF 0 EOL1-WeldelR 1585

-c f--<

~ WeldAX 1430 6= 10So "." I"~,~OOWeldAx1~

6= 28S"

12" J-,,~- -f

Plant Specific Results

24

Outline

• Plant specific analysisfeatures and inputs

• PRA

• TH • PFM

• Estimated yearly TWCF • Values • Distribution characteristics

• Dominant contributors to TWCF • Transients • Material features

• Applicability of these results beyond the 3 study plants • External events • Generalization to all PWRs

VG 53

Scope Considered PRA

• Initiators • LOCAs: small, medium, large • Transients: all types including support system initiators • SGTR • Steamline Breaks: small, large

• Types of accidents • Overcooling with lowering or otherwise controlled RCS

pressure • Overcooling with high RCS pressure • Overcooling with repressurization • RCS faults, secondary faults, and combinations of RCS &.

secondary faults • At full power and at hot zero power

• Operator Actions • Successes • Errors of omission • Acts of commission (procedu....riven) 54

25

Issues Important to Understanding the Results

• Numerous uncertainties were accounted for • Break size variation, • HPI flow and temperature variations, • Valve size openings, and • Timing of SRVreciosure

• Combinations of these uncertainties yield different TH profiles

• Representative cases were selected to depict all these possible TH profiles by the assignment of appropriate split fractions

VG S5

Issues Important to Understanding the Results (Continued)

For example, the original Palisades medium LOCA bin was subdivided into the following TH bins to represent the possible spectrum of TH profiles using the split fractions provide by UMD.

TH TH Case Description Split Case Fraction "1

62 20.32 cm (8 in) cold leg break. Winter conditions 0.35 assumed (HPI and LPI injection temp = 40 F, I\~~, ,~, ,I~i~r .~~~ - an c'\

63 14.37 cm (5.656 in) cold leg break. Winter 0.30 conditions assumed (HPI and LPI injection temp = An t: A ,I~._r ...~~ _ an c\

64 10.16 cm (4 in) surge line break. Summer 0.35 conditions assumed (HPI and LPI injection temp =

I "" r- A "" r-, , "I'" ~~ I

26

Plant-Specific TH Features Characteristic Oconee Beaver Valley Palisades

Plant Type B& W lowered loop Westinghouse CE design, 2x4 loops design, 2.4 loops, design, 3 loops OTSO

Core Power (MWth) 2568 2660 2530

RCP Trip Criteria all pumps assumed to 6P < 200 psid between the RCS one pump tripped in each loop trip when subcooling < and highest SO pressure (normal if PZR pressure < 1300psia. O.s"F containment conditions) All pumps tripped when

Llp < 375 psid (adverse subcooling < 25°F t . ·tion,)

HPI maximum flow 180 Ibm/sec 134.8 Ibm/sec 184 Ibm/sec

HPI shutoff head > 2600 Dsia > 2600 Dsia 1292 Dsia LPI maximum flow 1050 Ibm/sec 690.9 Ibm/sec 922 Ibm/sec

LPI shutoff head 214 psia 215 psia 218 psia

Accum liquid volume 2150 f( 3104.1 ft 4800 ft

Accum disch press 590 psia 648 psia 215 psia

PZR SRY Capacity 489,1831bmlhr 1,494,618Ibmlhr 690,0001bmlhr (total for 2 valves) (total for 3 valves) (total for 3 valves)

SG Waler Mass 40,000 Ibm (HFP) 118,760 Ibm (HFP) 142,138 Ibm (HFP) 11,000 Ibm (HZP) 160,470 Ibm (HZP) 210,759 Ibm (HZP)

SO SRY Capacity 13.0 Mlblhr 13.1 Mlb/hr 37.6Mlblhr (16 valves) (15 valves) (24 valves)

AFW maximum flow 1390 gpm (motor) 700 gpm (motor) 400 gpm (motor) (total) 1350 gpm (turbine) 700 gpm (lDrbine) 400 gpm (turbine)

VG 57

Plant-Specific TH Features (Cont.)

• Reactor vessel vent valves (B.W) • As in Oconee, valves connect upper plenum tIIIwncomel:

• LPI and Accumulator: • Beaver Valley and PalisadesLPI and accumulator

connections to each cold leg • Oconee- low pressure injection and core flood tank

(accumulator) discharge connected directly to the clowncomerabove the cold leg nozzle

• Palisades- low accumulator initial pressure

• HPJ flow characteristics • Oconee- about 30 percent more flow to the "A" loop

compared to the Bloop • Beaver Valley and Palisadesequal flow to all loops

27

Plant Specific PFM Features

• Cladding • Oconee: Single later clae»

eire. surface brea ing cracks

• Beaver valier. & Palisades Multi layer cae» no surface breaking cracks

VG 59

TWCF Estimates 1.E-05 .-------------,

>U I: 1.E-06 CIl ::I ~ 1.E-07

~ m f.E-OB 1:>~ g' 1.E-09

5 ~ 1.E-10 = 8~ ~ 1.E-11

.r:. g 1.E-12

~ 1.E-13... f.E-f4 .f-..~+-"~+-'~_t_'~'_t_'~'-l

o 200 400 600 BOO fOOO

EFPY [years]

--0-- Oconee • 5% -0- Oeon.. - Median

-4- Oconee • 95% ..... Ocon... M.an

_B••var·S'1iI -n- Beaver· Madian

-A-B..ver-95% -.- Beaver - Mean

___ Palis.d•• • 5% -0- Palisade•• Madian

~ Palisades - 95'/0 ... Pali.ades· Me.n

• All plants have plantspecific • Dimensions • Weld I plate placement • Fluence maps • Chemistry • Transition temperature

~, (I ~ (\\\/ \J \) \.) ~ -- ..

r------.- , , ..............•....• ' i

If-- --+-------+~------

I .. ·'."·._.'.'l

.... t" I 1

[1 II i ,

• OVer reall8tlc operatlonailifetlm., the e8t1mated TWeF fOr theM plants -yelY small

• Valu. range from 1xlG to 5xlO' • Two of these plants are among the

1£-05 ..-----------~

>g 1.E-06 +-------------1 CIl ::I CJ' f.E-07 -l.---~:a:::::::==;;::;o;o_:~e ~

~ m 1.E-OB {---------:",MO<::::=-=---1 I: >~ g' f.E-09 +---....~.....-===--=l(:)'" .-e f.E-fO +---_'0-....::"-----,..,:;;.....,£>.---1 = 8~ ~ f.E-11 +-----...,...,"-----..,L-----1

'§, f.E-12 +-----1----------1 ::>

~ f.E-f3 +-----1----------1 ... 1.E-14 +-~-+-~-+-~-+-~-+-~--l

o 20 40 60 80 100

ceDV r ............1

---tr-Beaver - 5% -trSe.va,. Median

--tr-Beaver . 9~J. ..... Seaver· Mean

....a-P.liaad... 5% ~Pali..d•• ~ Median

_Palisades - 95% ... Pali.ad•• · Mean

28

Characteristics of these TV CF Distributions

• Skewed: the 9!ih percentile • Brc ~: > 3 orders of and mean roughly coincide mal itude separate g. and

9sth ercentiles• ... because, the physical nature of cleavage fracture produces • ••. f all the same reasons finite minimum toughness list under "skewed" values • Dis butions narrow as plant

• ThereforeJ.Pr (init or fail) can ope ting time: because be, and onen is, zero mat ialembrittle$ mitigating

(or iminating) zere• However, sometimes (rarelypr con r butors to the TWCF(init or fail) is large • Severe transients, AND • Large flaws, AND ~:: It H.,::.:.• High embrittlement, AND

• These factors produc15kewed .s II t' ' TWCF distributions c

~ 0.50 .., 'l:

o 0.25

::J

L lC~ 0.00 u.. ' ........__ '0.

fE-11 fE-09 fE-07 fE.{)5 fE-II3

VG 75 Thru·Wall Cracking Frequency, TWCF

Dominant Transients Overview • LOCAs dominant contributor to risk • Stuck open valves also a contributor in law pw

due to plant design features •

fM,..------------, '--=:::::=:_::'.'.-~::~~-:::;:::=::::'.:::::::-.-~':~~~-I'J

-,-0--: s..o....... I JI:I:ft:..... l--""~lIftlfl I;I __~~(I ~,flI~ I l__~t>_·'_.. I!V--_· ...o _ ........ ! I I 1-~"fJ !

i

\~ !

----_'1

29

Contributors to Through-Wall Cracking Frequency

• TWCF is the product of • The initiating event

freguency, IEF, (){axis),and

• The conditional probability of failure,CPF, (V-axIs)

• The contribution of IEF and CPF to the through!Nail crac~ing frequency IS approximately"balanced'"

• All but two of the dominant transient categories havelEFsand CPFsthat are within about±l order of magnitude

VG 77

I.E.{)2 ,..----------.,===;;,

l!! .:! ;f I.E'{)3

'0 ~ :B I.E-M co .c e ll. I.E'{)5 iii c o ~ I.E.{)6 c o () o

I.E.{)T +,o===r<-----,-----,,---...,..---4 I.E-OT I.E'{)6 I.E-05 I.E-04 I.E-03 I.E-02

Initiating Event Frequency

• Oconee - LOCA (Pipe Break)

II Oconee - Stuck Open Valves, Primary Side

• Beaver - LOCA (Pipe Break)

X Beaver· Stuck Open Valves, Primary Side

X Beaver - MSLB

• Palisades - LOCA (Pipe Break)

+ Palisades· Stuck Open Valves, Primary Side

A Palisades - MSLB

o Palisades· Stuck Open Valves, Secondary Side

Source of Uncertainty in Dominant Transients -7 LOCAs ~

100

• LOCAs imp.ortant in all 3 p'lants, dominate in Palisades anCi Beaver

• Relative contribution typicallyincreases (or remains approximately constant al a high value) as EFPV increases ~

• About 3 orders of magnitude ~ 60 uncertainty inTWCFscfriven by: • 2 orders of magnitude come from ,g

uncertainty in LOCA frequencies C (reflect latest NRC expert judgments) :8 40 propagated thru the analyses .E

• T-H uncertainty is handled by different-;: "bins" each representing different 1: LOCA sizes; Within any bin- no 'FH <3 uncertainty (small) 20

• PFM uncertainties account for ';Ie remainder of uncertainty

./ 1 order of magnitUde: R;IlTbiasadjustment

./ 1 order of magnitude: flaw distribution • Operator actions do not playa key role

_.,:0 .

1 • . .. . .. 0+--.,....--.,....---r--.,....----1

o 200 400 600 800 1000

EFPY [years] 78

VG 78

30

Large Break LOCA Results Comparison

600 539 3000 207

500 5J3 2500 172

I 400 478_ 2000 138

ri:o. §, iii~ 0.."

;~ :300 l!' 1500 103l!'422i ~ ~

.~ .. ~ ~E'~ 0.. 0..f- 3Q6+200 1000 6.9

/"-

W ~--

100 311 500 34

0 o0 :JOOO 6000 9000 , 2000 15001T' o 0 3000 6000 9000 '2000 150& Time ($) nm<tiSI

-- Oc:onnCase 156-16"holltQ~ICPF. 2433E-001 -- Beaver Valley Case 009 - 16" hot log br'eaJ< (CPF. 2.29~E-OO} -- PahlitJW1pR.I.,:,,~p040 _ 'n" hJ'l4 ~. brpRIc ~r.PF .. 1 ~, F_n?~

VG '9

Medium-Break LOCA Results Comparison

207

2500 17.2

138

l e 103J

6.9

3C

0 0 3000 6000 9000 12000 150~ Ti"'" (oj

Beaver Valley Case 007 - 8" SUIlJ" line br'eaJ< (VFT4720s) PalIsade. ease 062 - 8' coid leg bresk. wi_ (VFT• 7:20s1

veM

31

Medium-Break LOCA Results Comparison

600 539 207

1 500 \ 533 2500 17.2

l'., "::i

400 . , ~78_

[~ , ~ ~ I

» 11 .. i .3 : I :; ! t.. 300 422e I!! 1500 !

t E ~

1: .~....-.",.. ffiJ ,'I' ~ ~ 1

,1.\1.'~ \ .... .a.....-.h.!..!,;j..._ .. 8. I

I

200 HlOO 6.9366....,. .,-

Q. l :'

I " .1 ,'r~ ~ 1

100 311 34 600 J,\'i"

\\.t; ..

oa ~'-;~~o"'-'~~';;~;~;o£O0 0 :lOOO 6000 9000 12000 I 500~ Time(sl Time (61

... Palisade. Case 003 - 5.666" o:>d leg break. ""'ler {YFT~ 1500s) P;liU:;AJ1fl1~ ('.$11':11' M.4 _ 4' ~1J,!!i=I tin,;. hrA::!IIk. j:jIJnvnRr ~VFT... ,,~)

VG81

Small-Break LOCA Results Comparison 600 589 :lOOO 207

I 500 533 2500 17.2

478_ 2000 138 ~ ¥ ~ 1r ! ~

422i! ! 1500 103 ~ ffi8. ..E ~ ~ Q. Q.

366.... 1000 6.9

100 311 3~

0 0 :lOOO 6000 9000 12000 150~ aor :lOOO 6000 9000 12000 150£a Ti"",(sl Time (s'

-- Beaver Valtey Cese 056 - .- surge line break ('1FT.1050&) PalISade. Case 058 • 4' cold teg braak. wlnt>!r lXlf1dI'ian. ('1FT.27008)

Hr. .,

32

Comparison of Dominant LOCA Transients

Oconee Beaver Palisades •.0,- ---, ..-,"-- ---mJ-.--.--,,--------------, r--.-- .... ..

-:)-156 -16-in. hot leG -0-7: 8-10. Surg_line Break -0-58: 4w cold legWinler

3.5 -<)-160· 5.&-in. ourge -+-9: 16-io. HotLeg Br..k -<>-59:4" cold '-V SlIlIIlTlllr

........60: 2- SUIgll Leu Winter

3.0 ...... ._"'ro". .......'A

"".7""""'.. WI"w ;" '1'"" ~ 2.5 64:4·.u~Summ8r ~J ~ J ~

11 •5 '

J I t.O

0.5

0.0 .~~---=;:::::::::;::'-----l"'''''''--~-~-~------l;'----~-~--~-200 100 ·100 ·20.., 100 "0f).2f#JO tOO -tOO.200

T • RTNDT fF] T. RT NOT ["F]T· RTNDT ["F]

Flaw size & location, and embrittlement constant in all three analyses. VG 83

Source of Uncertainty in Dominant Transients -7 Stuck Open SRVs that Re-Close Later ~

• Stuclt-open SRV I recl05ure type 100

scenarios an important class of transients for Oconee only

LL 80• Relative contribution lowers as EFPY increases ~

• Important for Oconee due to greater b tendency to decouple RCS from secondai 60 and less heat addition from steam >generators into the RCS during event in .s saw plants l:

• Key uncertainties in this type of transien~ 40 have been addressed quantitatively :9

" Degree of valve opening ~ -,. Modeled by • split fraction for fraction of valve 0

:::~:~: ~i~~:I~~I~~:; to PTS assuming any size (J 20

../ When valve recloses ';;;!. ,. Modeled by two discreet models (bin-i reclosureat

3000 sec keclosure at 6000 sec with 5850 probability

0+-".......--r-----r--..---,-----...1" How fast operator controls ReS PJlressurization , Modeled by different times and associated o 200 400 600 800 1000

probabilities with uncertainties for operator actions. 84 Note: different probabilities used across 3 plants; no EFPY [years]considerable credit for success.

33

Stuck-Open Primary SRV Results Comparison 000 589 = 207

500 533 2$(X) 172

, I,._ 400 , 478.~ 2000 ~ 138

~~ lO. .~ 'I ~ ,~ •, , ~ ~ :\ e

Il:~ 300 •, \ \

r'~;:'- 422~ 1500 1'\ '03 ~ \ I -- 8. ~ } ),,- .~ ,t]

;. \ \

\ " ..... ~ <l: t- 200 ~,, " ... 1 ,- 3001- 1000

I :1 It

69 1

I II! , ,. Ii,

......,..,... _1'100 " 311 500 :~ 34,-"'~:::'::~

0 ~ 0 00 0 = 6000 9000 I roOD 15000 0 JOOO 6000 9000 lroOO 15000

Time (s.1 Time (!I..I

-- Oconee Case 109 - RTT WI1 SO pzr SRV (reel@ 6000S1. no HP, WO"'JI'Il (Vl'T_7t4O$) --' Oconee Ca~ 113 - Rn WI' SO Pl' SRV (reel@6000S).HPltIltOlll1ng(10mindelay)(VH.714O$) -- - Ocon•• Case 122 - Rnwll SO pzr SRV ([email protected] (10 min dolay)(VFT..o9ros) ~ ~- ~liA9dPIt CRee 0fi..."1 ~ RIT w.'1 ~O Pzr ~RV (r~l @ Allt'\t)s~. H7P, no HPJ thmrttinQ (VFT_n:~~

VG 85

Stuck-Open Primary SRV Results Comparison

= 207

G. 11.2""1- ",',..rt*u 'M , 'I Ilji, ... ~''''''''. 138ft

£

\ 103 i I tI:

1000

100 - -. 311 \/....::~

...... _- ... _ I~~ ~____

o 2S5 o 00 o = 6000 9000 1roOO 15000 o = 6000 9000 lroOO 15000

nme(s) nme(s)

•__• Ocon•• c .... 124 - RTT WI' so pzr SRY (rael@3OOO!».HZP. HPlmrcm,ng (10 min delay) (VFT-43IlO!I) - _. Seav.r Valley Case 097 • RTT wll SO pzr SRY (rael@3OOO!>1.HlP.nGHPIIh'Gnling(YfT-2490Sj

He AA

34

mailto:rael@3OOO!>1.HlP.nGHPIIh'Gnling(YfT-2490Sj

mailto:rael@3OOO!�.HZP

mailto:[email protected]

mailto:reel@6000S).HPltIltOlll1ng(10mindelay)(VH.714O

Comparison of Dominant SO Primary

Oconee Beaver Palisades 4.0 r---------, ,----------,

...o-109-S0SRVgFP, 100 1mmin 1"Io000Ioe., no thrott..

3.5 __1f3. SO SRV@FP, 100 min re·clo••. wi throttlo

-6--115· SO SRV. FP, 50 __66: SO P-5RV !'KIa... -'tor 100 min. ...0-97: RTm with. Stuck Open SRV, r..

clos•• .t 3000 see. 3.0 ",in ro-elo.., wi ttlrottle

~: ...o-:~~·re~,::~~::ni:O ~i 2.5

:!!.

!2.0 ,z j 1.5

J 1.0

0.5

100 -100 ·20l10 100 ·100 -ZOO ~ 100 .100 .2Ot)

T· RTHOT [OF] T· RTHOT ['FJ T. RTHOT ["FJ

Flaw size & location, and embrittlement constant in all three analyses. VG87

Non-Dominant Transients -7 Main Steam Line Break? ~

100

• Main Steamline Breaks and other secondary

II. 80

faults (stucltopen ~ valves, overfeeds, etc.) >

are relatively ~ 60

unimportant oc: o

• Why? generally: :s 40 ,/ Binning not as gross as in eartier work ~

';:(current work separates large breaks 'Efrom small breaks from vahrepening o scenarios so there is less conservatism (,) 20 compared with earlier studies) f!!.

,( Not as severe a transient as a LOCA ,/ Realistic credit for operator actions

including uncertainties on human action o .m'i~I:::::=:::::;=:IiL....,....----.--1 probabilities o 200 400 600 800 1000

• Note: Uncertainties I judgments would EFPY [years]have to be sign~fic~ntly different before

35

Stuck-Open Secondary SRV Results

fooo

100 ~1 34

00o0 3000 5000 9000 , 2000 ,50Jirl 3000 6000 9000 '2000 '5000 Time(.i Time (E.~

VG 89

MSLB Results

600 589 207

500 533

,_ 400 476_ ~!:. ~ ..'.'. 300:s j

=1'S

.:+v''i ',. ..E ;, 3&;....200

100 311 3,4

0 0 3000 6000 9000 12000 150Jr' 0 0 3000 6000 9000 12000'50£0 Time (s) Time (s,

Palisades Gase 054 - MSLB ..''''SlY tailure, no HPllhrotillng. no AFW isolabon (VFT.""xo)

VG go

36

Comparison of Dominant Secondary Side

Oconee Beaver Palisades ~o r--------------,

.......... '02:MSLB lit full poW.... Jl,il1l.-'~ continu•• for 30 minlltes and -4-19: SO ADV, HPI not throttled

ope.-.l:or hold HSI open for 303.' minul.. -tr-52: SO .lOV & both MSlVs don't close

~ 103: MSLB at HZP. AUI-'-t continues fur 30 minutes Ind operwtor hold HSI open for 30 minute.

-'-55: 2 so AOVs, flw rrom 2 AVV Feed pumps

]'.5

J '.0

0.' lIDdO!-~~~~~=~~,....:::....~-~-~----< 200 100 ·100 -200 100 '00 ·'00

T • RTNOT [oF) T· RTNDT [oF)

VG91 Flaw size & location, and embrittlement constant in all three analyses.

Non -Dominant Transients -7 Main Steam Line Break?

2.0

• Not as severe ......a transient as ...

a LOCA ei

.51.5 «

~ ......

.., '" ~0.5 a... ~

0.0 200 100 o -100 ·200 -300

T· RT NOT [OF] 7: a-in. Sur e Line Break

......-9: 16·10. Hot Leg Break

-0-- 56: 4-in. Surge Line Break

..........-f02:MSLB at full power. Aux-f••d continues for 30 minute. and operator hOld HSI open for 30 minutes

-0--103: MSLB at HZP. Aux~f.ed continues for 30 minute. and operator hold HSI open for 30 minutes

~104: MSLB at full power. Aux·feed continues for 30 minutes and 37

Dominant Material Contributors

• Axial weld cracks dominate TWCF (-I04Vo) • Axla. weld Rl.m or • Plate R~

• Circumferential weld cracks playa minor role in TWCF .104Vo) • Cire. weld Rlim or • Plate R~ • Forging R'ta

• Cracks In plates and forgings too small to playa role

VG95

100

Yo 0

~ Axial. >;:

95 Welds III

~ ,g c 0 ;:: ;:,

90

,g".:.. c 0 0 ';Ie. 85

0 200 400 600 BOO 1000

EFPY [years]

Summary

i~__ ••l.IInl,l;tClli:;aill;.:~::adu.I,&:":''''~[;oiII:I&:;u[nll4:!Ion.&l1TI&".:IIil:IIIl~IIin:..n_t_I_Y_P_T_S_TW

• Operations • LOCAs and stucltopen

valves on primary side dominate PTS challenge

• Secondary side breaks insignificant contributors

• Holding all material factors constant, operational challenge reasonably consistent between 3 plants analyzed

'" Probability of challengeoccurring

'" Fracture challengeassuming event occurs

• Materials • (Nearly) all weld flaws

occur on weld fusion line • Axial weld flaws dominate

through wall crackingfrequency

'" Axial weld toughness '" Plate toughness

• Circ. weld flaws make minor contribution to through wall crackingfrequency

'" eire. weld toughness '" Plate toughness '" Forging toughness

• Flaws in large regions of plate" forging remote from welds too small to matter

38

Generic

Applicability of Results

Overview

• Consideration of external initiating events

• Generalization of 3 planl5pecific analyses to all PWRs

39

Impact of External Events

• External events (e.g., fires, earthquakes, floods) can also cause overcooling events

• Actual experience indicates a small fraction of significant overcooling events involve external events (1 or 2 out of over 100 events)

• Performed general bounding analyses to determine worst-case contributions to TWCF (e.g., no operator credit)

• Findings: • Realistically: Not quantified, but judged IRt be a

significant contributor tcJWCFsas compared to internal events

• Bounding results: External events could result in similar TWCFsas for internal events

99

VGOO

Example of External Event Approach

• Internal scenario #1 description • Small LOCA

• Corresponding external event scenario • Seismic-induced pipe break only external event

identified ./ 0.3 ~ HCLF assumed (corresponds to 0.5 g median

fragility) with uncertainty

• Based on above fragility information &. seismic hazard inputs, determined seismiCnduced small LOCA frequency for two sites • H. B. Robinsont.1E-4/yr (mean) • Diablo Canyon5.0E-4/yr (mean)

• Note: No accounting for seismic effect on HPI (assumed to work) &. no credit for operator mitigating actions

VG 100

40

Example of External Event Approach (cont'd)

• Internal scenario #2 description • Reactor trip with single stucllpen PORV

• Corresponding external event scenarios • Seismic-induced PORVopening (e.g., relay chatter)

../ 0.3 ~ HCLF assumed (corresponds to 0.5 g median fragility) with uncertainty

• Fir&induced PORV opening (e.g., hot short) • Based on above fragility information and seismic hazard

inputs, determined seismiCnduced PORV open scenario frequency for two sites • H. B. Robinsonl.1E-4/yr (mean) • Diablo Canyon5.0E-4/yr (mean)

• Note: No accounting for seismic effect on HPI (assumed to work) a. no credit for operator mitigating actions.

VG 101

Example of External Event Approach (cont'd)

• Based on 2E2/yr fire frequency (Aux Bldg electrical cabinets experience), 0.5 hot short p'robability, a. 0.1 factor to affectspeciffc cabinet/cirCUit of concern =H 3/yr fire-induced PORV open scenario

• Note: no accounting for operator actions such as to close valve/block valve, or other mitigating actions.

• Sum of seismic and firtiinduced scenario frequencies is <2E-3/yr.

vc '"'

41

Resulting Comparison of Internal V5. External Event TWCFs

(External Event TWCFs.$ highest Internal Event TWCFs)

Scenario Internal Internal Internal External External External Event Event CPFs Event Event EventCPFs Event

Frequencies @S-100 TWCFs Frequencies @ S-1 00 TWCFs

VG 103

Generalization of Results to Other PWRs

• Ongoing • Looking at 5 other PWRs that are among the most

embrittled plants

• Approach: • Compare plant design and operational features that

matter most to the 3 plants analyzed • Qualitatively judge potential impact on PTS results

based on these comparisons • Assuming LOCAs should still dominate, results

should be similar since frequencies would not change, T-H responses should be similar to extent plant features are similar, anGPFsshould not be drastically different from plants analyzed (Beaver Valley and Palisades are also among the more embrittled plants) 104

,~'''''

42

Plants Covered In Generalization Step Tol.rance to • MOlt Embrlttfed In'NPT(.) + lIT.ell.tlon'I.nt rume Shift .t 40 _ ... 'P'lI'TSCh.II.n•• M.hirl.1

.::q~ 2 BEAVER VALLEY 1 PlATE 194• Plants ranked in ~ui

3

1

8£ ii~terms of u", ;i 5 PAUSADES 119

: e 7irradiated R"DT + ~:Ei 8 ~ PLATE 17.!i • S9E ~Eason e'" 10 WATTS BAR 1 FORGING 164

U 11 ST. LUCIE 1 AXIAL WELD 164 ~ ~ 12 SURRY 1 AXIAL WELD 163embrittlement ~.l! 13 INDIAN POINT 2 PLATE 162

14 GINNA FORGING 161shift at 32EFPY. ~~ IS POINT BEACH 1 AXIAL WELD 159 16 FARlEY 2 PLATE 158'~i 17 MCGUIRE 1 AXIAL WELD 158

OCONEE 1 157 19 NORm ANNA 2 FORGING 155

j2s 20 SHEARON HARRIS PLATE 153

• eire. welds NOT H 18 AXIAL WELD

considered in "i

22 COOK 2 PLATE 152

'Il 21 NORTH ANNA 1 FORGING 153

~ 23 SALEM 2 AXIAL WELD 148

I~ 24 CRVSTAl RIVER 3 AXIAL WELD 141ranking

Bj1.

!~

25 CALVERTCLIF 2 PLA 139 'E"! 26 ROBINSON 2 PLATE 138

27 COOK 1 AXIAL WELD 138 28 FARLEY 2 PLATE 133

~~ 29 FARLEY 1 PLATE 133

~ .~ 30 ARKANSAS NUCLEAR 1 AXIAL WELD 129

Notes:

VG 105 IifIIiiBII

General Information Categories Examined in General ization Step

• Secondary Breaches

• Secondary Overfeed

• LOCA Related

• PORV and SRV Related

• Feed and Bleed Related

vr. 'cv;

43

Secondary Breaches

• Number ofMSIVs • Isolation capability with regard to other paths

(e.g.,ADV~ SDV~ TBVs) • Identification of procedures, steps, and location of

steps within procedures that ensure likelihood of early identification" isolation of faulted steam generators

• Operator training or procedural allowances that support early isolation of steam generators

• Location/size ofsteamlineflow restrictors • Key assumptions relative to MSLB analysis • AFW/MFW control durin§teamlinebreak (or

simirarl [e.g., auto isolate, selfimiting flow, manuar only control]

• Does turbinedriven pump isolate in MSLB?

VG 107

Secondary Overfeed

• Information on the feed (MFW and AFW/EFW) capabilities to the steam generators including inventory of water available to continue MFW or AFW/EFW

• Information on normal steam generatorinventory

• Information on possible feed temperatures for all feed sources (especially how cold they could be)

VG 108

44

LOCA Related

• Allowable range of safety injection water temperatures

• Information to estimatftcirculationwater temperature

• Safety injection water source size (i.e., inventory) • Safety injectioltlowrateversus LOCA break size • Charging, HPI, LPI shutoff heads • Actuation requ!rements for containment spray and flow

rate once running • Impact on tlPI, LPI, charging when sump switchover

occurs (which pumps on vs. off) • ~I'!y significan~ changes in flow rates going from

InJection toreclrculafiolf • Accumulator (SIT, CFT) discharge pressure

VG 109

PORV and SRV Related

• Number and sizes oPORVs ..SRVs, whether plant operates with PORV block valves normally shut, and if there are any auto operation features of tlAORVs

• Instrumentation available (e.~itCcousticmonitors, differential pressure, etc.) to Identify opeoRVsorSRVs and to notice if they havl!eclosed

• Procedure for addressing LOCAs resulting from stuck open PORVsor SRVs

• Procedures for addressing the suddseclosureof such valves including throttle/terminate SI guidance

• Training material associated with sudeclosureevents • Operating characteristics of charging whtpressurizer

level goes back high (e.g., stop, keep running?) • How manySRVs must open before likely initiation of

containment sprays?

45

Feed and Bleed Related

• Number of AFW/EFW pumps/flow paths versus minimum success criteria for adequate feed to the steam generators

• EOP criteria for initiation of feed and bleed • Number ofPORVs opened out of total availableJor

even SRVs if pumps can oper6RVs) when in fee and bleed mode

• Number of HPJ pumps used in feed and bleed and is actual flow rate equivalent to number of pumps (e.g., at BV, they attempt to use all pumps but design only allows 2 out of 3 pumps to be aligned for injection at anyone time)

VG 111

Summary

• External events • Contribution small relative to internal events

• Generalization • 5 plants selectee» highestembrittlement • Question for plants developed based on

understanding of important contributors developed so far

46

Reactor Vessel Failure Frequency

..ance Criteria

Reactor Vessel Failure Frequency (RVFF)

• RVFF criterion needed for two purposes:

• Sup'port definition of RPV emtirittlementcriterion

• Provide accep.tance criterion for safety analysis

• Current metric and criterion established in RG 1.154:

RVFF=TWCF

RVFF* =5 x l(J1i fry

• Limited sco~ activi" to revisit metric/criterion in light of recent risltinformed regulationinitiatives

VG 1t4

Gl )

=~ ~liil---"",--... :::l

~r IlUu..

&!e II~ i Scr~e~ing I ~u.. I T

.... > Vessel damage, age,

47

Task Activities

• Identification of options

• Scopingstudy of post!vessel failure accident progression

• Qualitative evaluation of technical issues

• Review of pilot plant calculations for T/H conditions

• Limited calculations

• Status reports and meetings

• 5ECY-02-0092 (5/10/02)

• ACRS (7/10/02), pUblic meetings (10/17/02; 1/31/03)

• Chapter 5, draft NUREG (12/31/02)

• Focus on accej)tabilitf => activities are largely independent of plan~specificstudies

VG 115

RVFF Acceptance Criteria Principles in Developing and Evaluating Options

• Consistency with intent of original rule

• Low risk level

• Low relative contribution

• Consistency with recent risinformedinitiatives

• Risk metrics

• Risk criteria

• Consideration of defenstin-depth

VG 116

48

RVFF Acceptance Criteria Options (SECY-02-0092)

• Definition of RVFF

• RVFF = f(PTSinduced RPV throughNall crack)

• RVFF = f(PTSinduced crack initiation)

• RVFF acceptance limits

• RVFF* =5 x lf1>/ry

• RVFF* =1 x If#/ry

• RVFF* = 1 x lf1'/ry

VG 117

Post-SECY Discussions

• BudQetinq process: focus effort on assessing RVFF for pilot plants

• ACRS Letter (7/18/02; ML0220406120)

• RVFF should be based on considerations of LERF (and not CDF)

• Current LERF surrogate goal is not proper starting point

.....source terms used to develop the current goal do not reflelld! tlir-oxidation phenomena that would be a likely outcome Of a PTS event."

• Options:

./ Develop acceptance criterion from prompt fatality safety goal

./ Use a frequenc:vbased approach to develop RVFF* to provide assurance that PTS-induced RPV failures are very unlikely

• ACRS' expectation: RVFF* will be substantially smaller than optiproposed in SEC¥02-0092

"" 11.

49

Definition of RVFF

It is appropriate to define RVFF as the frequency of through-wall cracks (TWCF)

• TWCF is a more direct indicator of risk than is the vessel cracking initiation frequency

• The current technology for predicting crack arrest is reasonably robust • Laboratoryscale experiments • Scaled-vessel experiments

VG 119

Seoping Study - Key Questions

• Is a PT5induced RPV failure likely to lead to melted fuel?

• Is a PTSinduced RPV failure likely to lead to a large, early release?

• Is the release spectrum (frequene,onsequence)for PT5-induced large, early releases significantly worse than that associated with risllignificant, non-PT5-induced scenarios?

50

Scoping Study - Approach

• Refine SEC¥02-0092 list of technical issues • Develop accident progression event tree (APET) to

support identification, representation and discussion of technical issues

• Evaluate current state of knowledge regarding technical issues

• Context for evaluations: • Focus onsilot plants; some consideration of plants

addresse in generalization task • Whether/how PTS changes accident progression

VG 121

Accident Progression Event Tree (APET)

initiates Circumferential

"" '"

51

l! 11 1 ! 1! Illll 1 2 31--Arrested al 4

ircumferential Small I-- s -O-lOin""'" d-----1---1 •£ ~

7

--------J

~1I __--1__-t==;I::::==I'°yibCii...... 't r--- 11

y y"lU·lo.:t.lI'Z. I---- 12 __ .•.•.•.•. w_··· 13.24 See Sequences 1-12

lt ---- ~ 27 Ircumferen- 213

29~ "' --111-_--J__-t--q::f==::::::==131- Design '"

Basis 32--r=:== 33~ - 34

",. : >Design J:'Tf----,L--t---t=:::::;~~~1 ",;~ ...

Inihates

. Circu~ferenti

~ ..._too Circum

~167 ferential r--- .. 188 ----,' erylarge 189

II.TCam 190="';;;".......'_',;,;,;OO;,;,;O';;:,')+-__,'1 ot-----Ir------ I ~~; 193lr-I

~ ---~ ~~ "Y"=y S r---l ~ ~ ~~

J>---~~·Y~'~"~yts~.~bu~tr:edlc~e~d:ma~gjntUd~e_-.L__L_~c~=:Jt;:;;:;;:;;:;;E:::~198"" n 199 ~ 200

Potential Sources of Dependence Between Top Events

• Plant systems • RPV movement • Fragments • Fuel movement

52

Plant Conditions at RPV Failure

• Power available, cooling systems running (injection mode)

• LOCA events: RCS cooling, depressurizing • MLOCA- RPV failure at "'15)0 min (40 EFPY) • LLOCA- RPV failure at "'510 min (40 EFPY)

• Stuck-open SRV events: RCS at SR'Ietpoint RPV failure at "'68120 min (40 EFPY)

VG 125


3.*" _ ... , 11'lF,iqud:lOh1 ----.- ClC::C.JnllrQ 109 ·:VF"'" .. 114&'1

3e.11 - Ooorlee n3 :...P .. ,., 4{ll, --" ~ '2'2 ~vr~ .. ,,9fHJ'1 --~'?4~fJF7.ustnJ

---.-~ 1S6 {,-'n .8'i&QI;2.*"

:~ 29...11.. i ,0 1.S.."

' .....11

se.,o

0 -aooo 0 3000 6000 9000 12000 lS000 Time ~.)

53

------- -----------------------------------

---


3<1+11

.. -, 111FilqudAdt31

----- Piili~~".v.lj"'F": .. 3J~)

- PahiD.:Hts. ~ .:'..r~ ,. ..~u)

- PBJ"lIOtl~ ~~ {"r • ~2:,1.~,1

- f'Ilihu'trt':'I!< $a :"r- • ~1':ln.,1

2.se.11

l~ ~2&+"

"',. 15e+11i

l5 1e- ... , 1

~

5e..,o

0 -aooo 0 Jooo 6000 9000 12000 15000

TimP ~~~

VG 127


3e+11

.... _. 717FI~ud"'dd

---'- Pdsaaas,6.2 ~'iF'7 .. 121ll) - P"hiil:la,63 :VP ,. 15001) - PII~I5~S'~)vr _ '92{1.~1

- P\lh..'.k", ~ lV~ • lu.liIl

"' ,~ 1,58.11

iu

1e..11

o 9000 12000 15000

54


- - • 2':?F :iq~ld :h:h1 ----.- a'.' DD7 I\lF'- .. iZtsl - 9 ... [W3 '\If I • ::lCesl - B'.' 05G I''''FT. ll}~c!;! _ FW f}Q7 I\lFl .. 2"~1t1",,1

j

~ 1.5&+11

L~ ------~---

o 3000 6000 9000 12000 15000 Ti~~q}

VG ll9

Pressure Differentials After RPV Failure

Plug-in from Dave Bessette

55

Key Sequences

VG 131

Ev·,1_1_] :_1_:_1_1 · -70

~15~tjt---¢==t~~~I69"72

-74 Y· 75~fiiCfi~iAiiUdt;;:=j~~~I7376 Y

r=--'iiiaiiiji77ndent 80

82""", 83OO-1.Q9..Q.!..n 84

86

f\+--~~"87 88b~~~I85

l--l.I(--r-1~EI!!

95

96 i~DepElndent~t~~t=j~~~~~~~~;."lIy Dependent

56

Observations

• Accident energetics are more benign than those of some other scenarios previously studied (e.g., HPME).

• Containment spray failure probability may decrease for PTS events. *

• Likelihood of fuel cooling dependent on reactor cavity design • Cavity flooding above top of fuel expected for some plants • For other plants, ECCS may not be sufficient to cool fuel

• Slowdown forces on RPV and internals likely to be the same order of magnitude or bounded by design basis LLOCA

*For some plants, this may be dependent on plant changes in response to GSI-191.

VGm

Scoping Study Conclusions

• The conditional probability of early fuel damage (given a PTS-induced RPV failure) appears to be

• Extremely small for plants with cavities likely to be flooded • Non-negligible for other plants

• The conditional probability of early containment failure and a large, early release (given a PT5-induced RPV failure) appears to be very small for all plants

• Should a PTS-induced large, early release occur, such a release may involve a large-scale air-oxidation source term

VG ''''

57

Implications for RVFF*

• RVFF* =1 x lfJ6/ry is consistent with philosophVof original PTS rule, with ACRS guidance, and with Safety Goal Policy Statement • Assures a low level of risk associated with PTS events • Assures small relative contribution to acceptable risk • More limiting with respect to core damage than RG

1.174/0ption 3 criterion for CDF • Consistent or conservative with respect QHOs

• Expectation: RP~mbrittlementlimitswill be established in a rislcinformed manner

VG 135

Summary Conclusions

• RVFF=TWCF • RVFF* =1 x lfJ6/ry • RVFF* should be compared against mean of plant

specific RVFF distribution

""""

58

PTS Screening Limit

Considerations

Outline

• PTS risk at likely operational lifetimes

• Operating challenge considerations

• Materials considerations

• A physically motivated embrittlement metric

VG ''''

59

~ 1.E-OSPTS Risk is Low ... --¢-- Oconee ~ I ; ... 1.E.06 --0- Palisade. ~ : --fr- a...... r

• Over realistic operational lifetimes, the estimated TWCF fOJ these plants 19ft small (from 1x1 to 5X1ctl)

• At the current screen~ criteria the yearly CF is '" approximately 1x1"

• Two of these plants an! .m!»~I. ~h~ .,:ost .

VG 111

Operational Considerations

All material factors held equal, the

severity of PTS chanen,eis remarkably

similar between thepllnts'.. -"

Gl Q. ~ 1.E-071/ \,u. en (.) 1.E-OB ~3: c 1-1.E-09

I';ftI

1.E-10

..!!! 1.E-OS .. ; ... 1.E..06 ~ ftI Gl Gl Q. >-1.E-07

~ \,~ en (.) tE-OB

~ ~1.E-09 ftI Gl

t.E-10::E

~~

/ ----Oft

0 2S 50 7S 100

EFPY [years]

fooconoe J 00 Palisades

!> Beaver I !> 0-f110

l> 00

0 100 200 300 400

., " Vi tea E 'C a. "

RG1.99 Rev-2 RTpTS rF]

~.'-;""'----" ... ". ....,........._... =_.._..

60

OconeeOperational Considerations

:.! u All material 0 ..... factors held equal, the

CDseverity of PTS 'a

challenge is (i

remarkably IIto Esimilar between i: A. ..the plants

studied lU 'a (iThe frequency of t-challenge is also ea

similar between 'ac plants 8

'H _eX .... ,.. VG 143 'H

T_RT.", ['F'] T_RT•• , [OF']

Axial Weld Flaws Dominate TWCF

• Axial weld cracks 100 ." I ,....., ..---. Idominate TWCF

(-IOItle) • AxI.1 weld R.or • PlateR.

• Circumferential weld cracks playa minor role In TWC:F~,.101tle) • Clrc. weld R'fer,or • PI.te...... • FOl'llnIRl.r

• Cracks In plates and . forgings toe sma" to

o 200 400 600 800 1000

EFPY [years]

61

Characteristics of a Physically-Motivated Embrittlement Metric

A causal relationship should exist between th. embrittlemenbnetric and TWCF ... so ...

• Axial weld I plate properties dominate the metric,

• eirc weld I forging I plate properties play minor role

• Relevant fluence is that along the welds

• Large regions of plate I forging remote from welds don't matter

VG 149

Embrittlement Met.,........,·OOI..----L--->O...................-.......-,

....---...j 3-10% Weight on eire. ~-1IIIIj

p - AWRTNDT*=O·Q+O·" Tl;:~~c;))]=-------,

Axial Weld , plate Reference Temperature Depends on

n!{Li .MAX [(RT,S'DT(U) +AJ;~V:)),(RT::r(u) +AJ;~WV::))]} T - =....,!;i=::l..-l _

p AW

n!L i

62

Suggested Embrittlement Metric & Si nificance

1.E-05 -r----.---..,-------.-----,•� VERY LOW predicted TWCF values suggest that revision of the PTS rule .. screening criteria is justified

•� A yearly RVFF limit of lxl06 events corresponds to a weightedRTNDT value (RTNDT*) of 290'F

•� SinceRTNDT * is about 90F less thanRTp7S, this suggests that a 88F to 110"F increase of the current 10CFR50.61 screening limit is possible

400

Results ......_that operation poalble for 10. 80 yea... without . *

RTNDT* Screening Limit for PTS

1.E-05 -r----.---..,-------.-----,

•� Margin onRTNDT* nelth!!r necessary nor appropriate

1.E-06 .t==+===1==iJ-=-~ •� Maximum material

uncertainties accounted for explicitly In FAVOR calculatlons- any plant -U. 1.E-07 +---+--~~-t-----1 state of knowledge will be better than we simulated ~

< l: 1.E-OB +---+---!'1t •� 2900F RT * limit pertains ca '-"MItIlJiDT II)�

only toRI NDT* estimated :E�from 1.E-09 +----j---I@~

•� RVID RTNDT(uj values •� Cu •� NI •� D

63

Significance�

Conelusions

• These analyses provide a technical basis to recommend revision of the PTS rule •� Two of the most embrittled plants in fleet have a

TWCF at or below 5x1CJ at end of license extension (60 years)

•� At the 10CFR50.61 Rlip'screening limits these plantshave a TWCF of 1xU'·'lVS. RG 1.154 at 5xU')

•� Analysis supports a revised screening limit of •� 290"F on a weighted RIDT value�

./ Axial welds&. plates dominate�

./ eirc welds and forgings minor contributors� • This limit is 80F to 1UJF higher than current�

10CFR50.61 limits on ~s

154

64

On-Going Activities

• RES activities •� Calvert cliffs • Generalization to all plants •� Sensitivity studies a. a more detailed examination of

current results • Favor Va.V •� External peer review of project •� Implications for operational limits (10CFR Appendix G)

• NRR activities •� RES Draft NUREG sent to NRR on 1-31-02 • NRR comments due by 331-03 • Decision to proceed with rulemaking?

VG155

65

,,~

Q)Q) ell

c c Ou. Q) al >. ~5

0 e e C', u E C'- 1ij C'. ~ ~ C'.

,,~ Cl

e� 0) 0) g .... 0 ~ EC'- E� '" u.O Ce

.><§ 'u; 0", .-e"O

0) e 01

.Q ~~ >.0. Q) ~~ w 5l~ U?:2~

.><e al- '- '" e m ,S ~ :J >. 0) 'u; Q) lO 'iii 0) 5 'iiiC o e� -eu -(6m (0) ~~ e_ E~:g Ocll::: ~tE! tT ~ '- U) e"~ rn ~ fI)al.!!! lO- 0)0 Q) I ...... 0'" alOO 1ll1llO al.!:::oI-~ ... x ~.~ _0 00 o Q. 0 ~oe..w 00 Ow :r:m mu. O.!!1 oms u.---l fJ38,2 ~tE! (/) woe.. ...Ja::e.. ---l«e..

Arrested at Circumferential Weld ... Beyond Circumferential Welds...

70 y' 71 72 Y Y

v" .., _� 73(-~'1~~O') ~;o I� : i 69

74 y' 75 76 Y Y

J� 77I� A~ .-J _ _. i 78 y' 79No� ..•-11'"---.-..-.-. 80 y y' Y

82 y' 83 84 y Y Y 85 86

_1O~~r6~0 in'ill : 81

y I 87

88 Y Y

90 y' 91

» Design -L------1===== 89

Basis� 92 Y Y 93Initiat~

Circumferential 95 Crack� lL -c===== 94 y'

y'96 Y Y 97

I 98 y' 1 99

100 Y Y~ y

V\� .----- ... Very Large� y'r-----{=====Complete (»1000 in')� ,.:.:R::.P.:.V__~ 116 .� 117IL..._-{===

115

":;:00 r----fl 118 Y y' Y ~ I Cavity 119

I 120<l-. }J 121 Y y' y <) J 122 .0 0� y'A ted .,.:.;Ro:.p~V __� 123....lr-----{=====

124NY! Circum· rres • • •� I l-----iC==== y125 Y Yferential� I LS:C~a:l:::vl!!:itvL.._--r---------- 126

127 ~O'Comolete • • •� yIL-----1C==== 128 Y Y

~.rP .J Jl,1

Meeting Minutes of the ACRS Materials & Metallurgy ... · Dr. William Shack, Vice Chairman of the ACRS Materials and Metallurgy Subcommittee, presiding, convened the meeting and stated

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