ML11157A010

STP Nuclear Operating Company, Licensee Handouts, 6/2/2011 Meeting Risk-Informed GSI-191 Closure Plan for GSI-191 Resolution

ML11157A010
Person / Time
Site:	South Texas
Issue date:	06/02/2011
From:	Grantom C South Texas
To:	Plant Licensing Branch IV
Singal, Balwant, 415-3016, NRR/DORL/LPL4
References
TAC ME5358, TAC ME5359
Download: ML11157A010 (107)

Text

RiskInformedGSI191Closure Plan RiskInformedGSI191Resolution Thursday,June2,2011 8:00a.m. 5:00p.m.

p PublicMeetingwithSTPNuclearOperatingCompany RickGrantom SouthTexasProject 6/2/11 Pre-Licensing Meeting No.1 1

RISK INFORMED GSI 191 CLOSURE ROADMAP 6/2/11 Pre-Licensing Meeting No.1 2

2011 GSI-191 Integrated Closure Plan 6/2/11 Pre-Licensing Meeting No.1

Conceptual Data and Computational Flow Diagram Risk Informed GSI-191 Resolution Thursday, June 2, 2011 8:00 a.m. - 5:00 p.m.

p Public Meeting with STP Nuclear Operating Company Ernie Kee South Texas Project 6/2/11 Pre-Licensing Meeting No.1 1

2 6/2/11 Pre-Licensing Meeting No.1

3 6/2/11 Pre-Licensing Meeting No.1

4 6/2/11 Pre-Licensing Meeting No.1

5 6/2/11 Pre-Licensing Meeting No.1

6 6/2/11 Pre-Licensing Meeting No.1

Licensing Strategy Licensing Strategy Risk Informed GSI-191 Resolution Thursday, June 2, 2011 8:00 a.m. - 5:00 p.m.

p Public Meeting with STP Nuclear Operating Company Jamie Paul South Texas Project 6/2/11 Pre-Licensing Meeting No.1 1

Licensing Strategy Request an exemption from certain requirements of 10 CFR § 50 46 10 CFR Part 50 Request an exemption from certain requirements of 10 CFR § 50.46, 10 CFR Part 50 Appendix A, and 10 CFR Part 50 Appendix K in order to use a risk-informed approach to resolve Generic Safety Issue (GSI) 191.

Specifically, STP requests a partial exemption from the provisions of 10 CFR § 50.46(b)(5), to the extent that it requires a calculation of long-term cooling of the reactor core for events that cumulatively have a very small risk (i.e., allow the risk informed approach to be an alternative calculation for determining events that have a very small risk, i.e., cumulatively fall with R

i III f R l

G id 1 174)

Region III of Regulatory Guide 1.174)

In determining which long-term cooling events that cumulatively fall with Region III of Regulatory Guide 1.174, request a partial exemption from 10 CFR § 50.46(a) and (c), to the extent that they require consideration of break sizes up to and including a double-ended t

f th l t

i i

th t

l t

t (RCS) rupture of the largest pipe in the reactor coolant system (RCS).

STP would consider break sizes up to and including a double-ended rupture of the larges pipe in the reactor coolant system but would further evaluate the risk contribution to CDF to determine the effects of debris accumulation on the performance of the sump during long-term cooling and to determine that the change in risk for those breaks cumulatively fall within Regions III of Regulatory Guide 1.174.

In addition, STP requests a partial exemption from the provisions of 10 CFR § 50.46(a)(1) and Appendix K to 10 CFR Part 50 to the extent they require use of an acceptable evaluation model to determine the effects of debris accumulation in the calculation long-term cooling pursuant to 10 CFR § 50.46(b)(5).

Allow the use of a risk-informed model as an acceptable alternate evaluation model to calculate debris generation debris Allow the use of a risk informed model as an acceptable alternate evaluation model to calculate debris generation, debris transport, sump performance, and in-vessel effects due to debris-related flow restrictions. This risk-informed model would be used for the determination of events that cumulatively fall within Region III and the calculation of long-term cooling for the remaining events.

An evaluation would be included with the exemption request providing the justifications for the ti d th t h i l

th d d i th l

l ti It ld f

ith exemption and the technical methods used in the calculations. It would conform with Regulatory Guide 1.174 requirements, including the provisions related to defense-in-depth, safety margins, and monitoring.

6/2/11 Pre-Licensing Meeting No.1 2

Containment Building CAD Model Risk Informed GSI-191 Resolution Thursday, June 2, 2011 8:00 a.m. - 5:00 p.m.

p Public Meeting with STP Nuclear Operating Company Timothy D. Sande 6/2/11 Pre-Licensing Meeting No.1 1

Section View Risk Informed GSI-191 Section View 6/2/11 Pre-Licensing Meeting No.1 2

Operating Deck Risk Informed GSI-191 Operating Deck 6/2/11 Pre-Licensing Meeting No.1 3

SG Compartment View 1 Risk Informed GSI-191 SG Compartment - View 1 6/2/11 Pre-Licensing Meeting No.1 4

SG Compartment View 2 Risk Informed GSI-191 SG Compartment - View 2 6/2/11 Pre-Licensing Meeting No.1 5

SG Compartment Floor Risk Informed GSI-191 SG Compartment Floor 6/2/11 Pre-Licensing Meeting No.1 6

Pool Elevation View 1 Risk Informed GSI-191 Pool Elevation - View 1 6/2/11 Pre-Licensing Meeting No.1 7

Pool Elevation View 2 Risk Informed GSI-191 Pool Elevation - View 2 6/2/11 Pre-Licensing Meeting No.1 8

Pool Elevation View 3 Risk Informed GSI-191 Pool Elevation - View 3 6/2/11 Pre-Licensing Meeting No.1 9

Piping and Insulation Risk Informed GSI-191 Piping and Insulation 6/2/11 Pre-Licensing Meeting No.1 10

RCS System Piping and Insulation Risk Informed GSI-191 RCS System Piping and Insulation 6/2/11 Pre-Licensing Meeting No.1 11

Insulation on Hangers and Valves Risk Informed GSI-191 Insulation on Hangers and Valves 6/2/11 Pre-Licensing Meeting No.1 12

Work Points for Hangers/Valves/Welds Risk Informed GSI-191 Work Points for Hangers/Valves/Welds 6/2/11 Pre-Licensing Meeting No.1 13

Work Points for Hangers/Valves/Welds Risk Informed GSI-191 Work Points for Hangers/Valves/Welds 6/2/11 Pre-Licensing Meeting No.1 14

Risk Informed GSI-191 PRA Modeling PRA Modeling Risk Informed GSI-191 Resolution Thursday, June 2, 2011 8:00 a.m. - 5:00 p.m.

Public Meeting with Public Meeting with STP Nuclear Operating Company Presented by David H. Johnson, Sc.D.

6/2/11 Pre-Licensing Meeting No.1 1

S:\\SharedFiles\\Local\\Pubs\\VUGRAFS\\DHJ\\STP June 2, 2011\\PRA Modeling - STP 6-2-11

Risk Informed GSI-191 Discussion of Status and Direction of PRA Modeling 6/2/11 Pre-Licensing Meeting No.1 2

Risk Informed GSI-191 Near Term PRA Activities Near Term PRA Activities

• Develop Expanded LOCA Event Trees p

p

- Detailed Sequence of Events Associated with Potential Sump Performance Phenomena

- Develop Necessary Logic Structures to Represent Develop Necessary Logic Structures to Represent Results from Other Team Members

- Support Uncertainty Calculation

• Integration of New Analyses

• Integration of New Analyses

- Utilize Analyses from Other Team Members

- Use PRA Framework to Assist in Technical Direction of Specialized Analyses of Specialized Analyses

• Use Existing PRA Model to Identify Potential Analysis Boundaries 6/2/11 Pre-Licensing Meeting No.1 y

3

Risk Informed GSI-191 LLOCA Event Sequence Diagram LLOCA Event Sequence Diagram 4

Risk Informed GSI-191 LLOCA Event Tree LLOCA Event Tree 5

Risk Informed GSI-191 SI Recirculation Fault Tree (Top)

SI Recirculation Fault Tree (Top) 6

Risk Informed GSI-191 Current and Future PRA Models Current and Future PRA Models Existing PRA

- Does Not Explicitly Address Sump Plugging Phenomena D

I l d A B i E (1E 5)

R U

il bili f

- Does Include A Basic Event (1E-5) to Represent Unavailability of Sump; Required for All Sequences Involving Recirculation

- At-Power Importance:

• FV 4 1E-03; RAW: 4 1E+02 [CDF]

FV 4.1E 03; RAW: 4.1E+02 [CDF]

• FV 2.4E-07; RAW: 1.0E+00 [LERF]

- Does Not Address In-Core Phenomena of Material Passing Strainer Current Effort

- New Model Will Depict Detailed Representation of Phenomena

• Aid in Investigation A

i t i D

t ti f Eff t

• Assist in Documentation of Effort

- Eventual Incorporation into PRA Model of Record

• Model May Be Higher Level, As Appropriate

• Could Result in Revised Initiator Groupings 6/2/11 Pre-Licensing Meeting No.1

• Could Result in Revised Initiator Groupings 7

Risk Informed GSI-191 Sensitivity Analyses y

y 8

Risk Informed GSI-191 Logic Changes Underway Logic Changes Underway

• Break Out Sump Blockage Logic Break Out Sump Blockage Logic

• Add Logic to Represent In-Core Downstream Phenomena Downstream Phenomena

• Add Logic to Represent Potential O

ti l St t

i Operational Strategies

• Differentiate LOCAs based on Potential Sump Impacts 6/2/11 Pre-Licensing Meeting No.1 9

Risk Informed GSI-191 Break out Sump Blockage Logic Break out Sump Blockage Logic

• Current Model Current Model

- Basic Event Representing Sump Blockage in Fault Trees with Other Causes of Failure

- The Single Basic Event Fails All Three Sumps

• Desired Model

- Break Out Sump Blockage Logic to Increase Visibility

- While Retaining Common Cause Failure Mode, Also Allow Individual and Pairs of Sumps to Fail Allow Individual and Pairs of Sumps to Fail

• Based on Debris Amount and Flow

• System Operation State 6/2/11 Pre-Licensing Meeting No.1 10

Risk Informed GSI-191 LARGE LOCA REACTOR POWER DECREASED CORE BECOMES SUBCRITICAL SOUTH TEXAS PROJECT LARGE BREAK LOCA EVENT SEQUENCE DIAGRAM CURRENT MODEL 4

SSPS ESFAS PZP CHI ACP DCP SI LHSI 6

CORE REMAINS SUBCRITICAL DUE TO INJECTING BORATED WATER IA 10 11 IB IC IA IB IC 12 RHR HX 13 STABLE CONDITION RA RB RC RX CORE BECOMES SUBCRITICAL DUE TO VOIDING. RCS PRESSURE BELOW LHSI SHUTOFF HEAD CCW RWST/ECCS COMMON PA PB PZ 3

ACCUMULATOR SAFETY INJECTION AI 2

INJECTION COMMON TO LHSI AND ACCUMULATOR 1

SI38A SI38B SI38C HOT LEG RECIRCULATION 15 H LEG FROM RWST ACP DCP 7

OR HI MANUAL SI OR MANUAL START HHSI CONTAINMENT SPRAY CS 8

5 M

WET LOSS OF SI PUMP NPSH CF CCW ACP DCP 14 RCFC CS WET DRY CH3 CS LT2 EARLY LOW PRESSURE ACP DCP 8

CONTAINMENT SPRAY

SSPS, SEQUENCER SS 9

LT2 LATE LOW PRESSURE WET SPRAY Q

COMMON WET DRY LT2 EARLY LOW PRESSURE 11

Risk Informed GSI-191 12

Risk Informed GSI-191 Add Logic to Represent Downstream Effect

• Current Model Current Model

- Fuel Damage Due to Material Passing Screens Not Included Screens Not Included

• Desired Model Include Scenarios Addressing In Core

- Include Scenarios Addressing In-Core Phenomena Due to Material Passing Strainer Will Assign Any Damage Scenarios to Unique

- Will Assign Any Damage Scenarios to Unique Damage States 6/2/11 Pre-Licensing Meeting No.1 13

Risk Informed GSI-191 14

Risk Informed GSI-191 Add Logic to Represent Potential g

p Operational Strategies

• Current Model

- Follows Current Procedures

• Desired Model

- Explicitly Represent Actions That Could Explicitly Represent Actions That Could Influence Amount of Debris Arriving at Individual Screens (e.g., Cycling RHR Train) 6/2/11 Pre-Licensing Meeting No.1 15

Risk Informed GSI-191 16

Risk Informed GSI-191 Differentiate LOCA Initiators Based on Differentiate LOCA Initiators Based on Potential for Sump Blockage

• Current Model

- Single Initiator Categories for Small

- Single Initiator Categories for Small (Isolatable and Nonisolatable), Medium and Large LOCAs g

• Desired Model

- Identify Explicitly Those Specific Break Identify Explicitly Those Specific Break Locations That Impact Debris Phenomena 6/2/11 Pre-Licensing Meeting No.1 17

Risk Informed GSI-191 18

Risk Informed GSI-191 Incorporating Results from Detailed A

l Analyses

• As results of detailed analyses become y

available, their insights will be reflected in PRA model

- Interpretation of results of detailed analyses to support scenario-p y

pp specific basic events and event tree branching likelihoods

- Potential changes to model structure possible

• Support scenario-specific basic event and event pp p

tree branching likelihoods

- E.g., likelihood that NPSH requirements are not met given specific break size and location

• Further refinements to model structure

- E.g., analyses may show scenario timing requires more detailed representation

- E.g., analyses may show more details needed to represent downstream effects 19

LOCA Initiating Event Frequencies and Uncertainties Risk Informed GSI-191 Resolution Thursday, June 2, 2011 8:00 a.m. - 5:00 p.m.

p Public Meeting with STP Nuclear Operating Company Karl N. Fleming KNF Consulting Services LLC Bengt O. Y. Lydell 6/2/11 Pre-Licensing Meeting No.1 1

Risk Informed GSI-191 Discussion Topics

• LOCA frequencies scope and objectives

• Technical approach

• Key inputs and outputs

• Interfaces with other GSI-191 tasks

• Interfaces with other GSI-191 tasks

• Issues and strategies for resolution 6/2/11 Pre-Licensing Meeting No.1 2

LOCA Frequency Project Team Risk Informed GSI-191 LOCA Frequency Project Team K

l Fl i

P id t KNFCS LLC Karl Fleming, President KNFCS LLC Former VP at PLG Inc. and ERIN Engineering and Research PLG project manager for initial STP PRA and early risk-informed applications Principal author of EPRI projects on RI-ISI, IFPRA, and EPRI pipe failure data j

Bengt Lydell, VP, Sr. Principal Consultant, Scandpower Inc.

Passive component reliability, IFPRA, RI-ISI, OECD/NEA Consultant NUREG-1829 Team Member P i i

l h

f EPRI j

RI ISI IFPRA d EPRI i

f il d

Principal author of EPRI projects on RI-ISI, IFPRA, and EPRI pipe failure data Fred Simonen, Sr. Principal Consultant, Scandpower Inc.

ASME Fellow Fracture mechanics structural integrity PTS RI-ISI Fracture mechanics, structural integrity, PTS, RI ISI NUREG-1829 and NUREG-1806 (PTS) Team Member Steve Gosselin, Sr. Principal Consultant, Scandpower Inc.

ASME Fellow Fracture Mechanics, structural integrity, RI-ISI, TLAA 3

6/2/11 Pre-Licensing Meeting No.1

Team Relevant Experience Risk Informed GSI-191 Team Relevant Experience I

l t i EPRI RI ISI Involvement in EPRI RI-ISI program Co-authors of EPRI RI-ISI Topical Report Developed Bayes methods for pipe failure rate development Developed piping failure rates for EPRI RI-ISI evaluations EPRI TR-111880 Developed Markov model for evaluating impact of inspections on rupture frequency Achieved NRC approval for these methods and databases for use in RI-ISI LARs Peformed 27 plant specific RI-ISI evaluations using EPRI RI-ISI method including risk impact evaluation for STP RI-ISI pilot study g

p p

y Developed piping system failure rates for IFPRA and HELB PRA for selected U.S. utilities Developed piping system failure rates and PRA procedures for IFPRA and HELB PRAs (EPRI Reports )

HELB PRAs (EPRI Reports )

Developed the technical basis for Division 2 of ASME Section XI for modular HTGRs Expert panel involvement in NUREG-1829 (LOCA Frequencies) and NUREG-1806 (PTS) and input to NUREG/CR-5750 4

6/2/11 Pre-Licensing Meeting No.1

LOCA Frequencies Team S

f W k

Risk Informed GSI-191 Scope of Work Incorporate insights from previous work on LOCA frequencies Characterize LOCA initiating events and there frequencies with respect to:

Characterize LOCA initiating events and there frequencies with respect to:

Specific components, materials, dimensions Specific locations Range of break sizes g

Degradation mechanisms and mitigation effectiveness Other break characteristics, e.g. speed Quantify both aleatory and epistemic uncertainties; augment with sensitivity studies Support interfaces with other parts of the GSI-191 evaluation LOCA initiating event frequencies for PRA modeling B

k h t

i ti f

l ti f d b i f ti Break characterization for evaluation of debris formation Participate in NRC workshops 5

6/2/11 Pre-Licensing Meeting No.1

LOCA Frequency Technical A

h Approach Utilize passive component reliability methods and data from RI-ISI technology Estimate STP specific failure rates vs. break size at each weld location based on simple models developed in EPRI RI-ISI; consider breaks size up to an including DEGB Inform estimates based on results of RI-ISI damage mechanism evaluation Utili M

k d l t l

t i

t f NDE l

t l

ti f

RI ISI Utilize Markov model to evaluate impact of NDE element selection from RI-ISI Aggregate as appropriate to support PRA and debris generation models; account for relatively large weld population for 4-Loop PWR with 3 train ECCS Utilize PIPExp database to help resolve uncertainties in failure rates Incorporate service experience through 2010 Improve estimates of component exposure Apply Bayes method from EPRI RI-ISI program Utilize information from NUREG-1829 and NUREG/CR-5750 in optimum manner p

Major source for quantification of epistemic uncertainty Provides basis for comparison of resulting LOCA initiating event frequencies Consider probabilistic fracture mechanics evaluation on selected locations as may be required may be required 6

6/2/11 Pre-Licensing Meeting No.1

Key Results from NUREG-1829 Key Results from NUREG 1829 7

6/2/11 Pre-Licensing Meeting No.1

LOCA IE Frequency Model 1 of 2

i ix i

x m

LOCA F

)

(

(1)

I F

R P

)

(

(2) ik ik x

k ik ix jx I

F R

P

)

(

(2)

Where:

)

(

x LOCA F

FrequencyofLOCAofsizex,perreactorcalendaryear;subjectto i t i

t i t l

l t d i M t C l

epistemicuncertaintycalculatedviaMonteCarlo

i m

Numberofpipeweldsoftypei; eachtypedeterminedbypipesize, weldtype,applicabledamagemechanisms,andinspectionstatus (leaktestandNDE);nouncertainty

ix

Frequencyofruptureofpipelocationjbelongingtocomponenttype ix

iwithbreaksizex,subjecttoepistemicuncertaintycalculatedvia MonteCarlo

ik

Failurerateperweldyearforpipecomponenttypeiduetofailure mechanismk;subjecttoepistemicuncertaintydeterminedbyRIISI Bayesmethod andEq. (3) y q ( )

)

(

ik x F R

P Conditionalprobabilityofruptureofsizexgivenfailure ofpipe componenttypeiduetodamagemechanismk;subjecttoepistemic uncertaintydeterminedviaexpertelicitation(NUREG1829)

ik I

Integritymanagementfactorforweldtypei andfailuremechanismk; subject to epistemic uncertainty determined by Monte Carlo and 8

6/2/11 Pre-Licensing Meeting No.1 subjecttoepistemicuncertainty determinedbyMonteCarloand MarkovModel

LOCA IE Frequency Model 2 of 2 q

y ik ik ik T

N f

n n

(3) i i

ik ik ik T

N f

ik n

Numberoffailuresinpipecomponent(i.e.weld)typeiduetofailure mechanismk,verylittleepistemicuncertainty C

t l ti f

ld f t i

tibl t

ik

Componentexposurepopulationforweldsoftypeisusceptibleto failuremechanismk,subjecttoepistemicuncertaintydeterminedby expertopinion

ikf

Estimateofthefractionofthecomponentexposurepopulationfor weld type i that is susceptible to failure mechanism k subject to weldtypeithatissusceptibletofailuremechanismk,subjectto epistemicuncertainty;estimatedfromresultsofRIISIforpopulation ofplantsandexpertopinion

i N

Estimateoftheaveragenumberofpipeweldsoftypeiperreactorin the applicable reactor years exposure for the data collection; subject theapplicablereactoryearsexposureforthedatacollection;subject toepistemicuncertainty;estimatedfromresultsofRIISIfor populationofplantsandexpertopinion

iT

Totalnumberofreactoryearsexposureforthedatacollectionfor componenttypei;littleornouncertainty 9

6/2/11 Pre-Licensing Meeting No.1

Estimation of Failure Rates ()

Estimation of Failure Rates ()

• Source data comes from Scandpowers PIPExp p

p comprehensive piping system database

• Bayes method for uncertainty treatment developed during EPRI RI-ISI program and approved by NRC g

p g

pp y

• Uncertainties due to scarcity of data, component exposure, and modeling accounted for

• Method has been refined and applied over many Method has been refined and applied over many projects for LWRs and HTGRs since 1997

• Methods and database recently applied to develop EPRI pipe failure data handbook for IFPRA and p p HELB PRA (EPRI 1021086)

• Methods being refined to address aging as part of DOE/INL Risk-Informed Safety Margin y

g Characterization Project 10 6/2/11 Pre-Licensing Meeting No.1

Volume of Available Data Records f

PWR Cl 1

d 2 Pi i for PWR Class 1 and 2 Piping 6/2/11 Pre-Licensing Meeting No.1 11

Bayes Methodology for Failure Rates Bayes Methodology for Failure Rates Number of Leaks Number of Leaks Number of Leaks And Ruptures for Specific System, Pipe Size, and Damage Mechanism Three Estimates of Component population Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p=.25)

Three Estimates of DM Susceptibility Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p=.25)

Lognormal Distribution Based on engineering Judgment and weld and DM susceptibility Estimates from RISI Number of Leaks And Ruptures for Specific System, Pipe Size, and Damage Mechanism Three Estimates of Component population Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p=.25)

Three Estimates of DM Susceptibility Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p=.25)

Lognormal Distribution Based on engineering Judgment and weld and DM susceptibility Estimates from RISI 6 00E-03 8.00E-03 1.00E-02 1.20E-02 1.40E-02 bili ty Density 600E-03 8.00E-03 1.00E-02 1.20E-02 1.40E-02 abili ty De nsity 8.00E-03 1.00E-02 1.20E-02 1.40E-02 1.60E-02 abili ty De nsity Bayes Update for Three Combinations of Population and DM Susceptibility 6 00E-03 8.00E-03 1.00E-02 1.20E-02 1.40E-02 bili ty Density 600E-03 8.00E-03 1.00E-02 1.20E-02 1.40E-02 600E-03 8.00E-03 1.00E-02 1.20E-02 1.40E-02 abili ty De nsity 8.00E-03 1.00E-02 1.20E-02 1.40E-02 1.60E-02 abili ty De nsity 8.00E-03 1.00E-02 1.20E-02 1.40E-02 1.60E-02 abili ty De nsity Bayes Update for Three Combinations of Population and DM Susceptibility 0.00E+00 2.00E-03 4.00E-03 6.00E 03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probab Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E 03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Rate Proba Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Proba Prior Posterior Forecast: Rupture Rate Forecast CE RC S TF F AILURES Lower-Lower Update (p=.1)

Best-Best Update (p=.80)

Upper-Upper Update (p=.1)

, Failure Frequency P(RF), Prob.Rupture Given Failure

, Rupture Frequency Bayes Posterior Weighting Operation Generic Prior Distribution 0.00E+00 2.00E-03 4.00E-03 6.00E 03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probab Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E 03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Rate 0.00E+00 2.00E-03 4.00E-03 6.00E 03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Rate Proba Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Proba Prior Posterior Forecast: Rupture Rate 0.00E+00 2.00E-03 4.00E-03 6.00E-03 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Proba Prior Posterior Forecast: Rupture Rate Forecast CE RC S TF F AILURES Forecast CE RC S TF F AILURES Lower-Lower Update (p=.1)

Best-Best Update (p=.80)

Upper-Upper Update (p=.1)

, Failure Frequency P(RF), Prob.Rupture Given Failure

, Rupture Frequency Bayes Posterior Weighting Operation Generic Prior Distribution Generic Prior Distribution Frequency Chart Ruptures per Suscepti bl e Weld Year Mean = 7.26E-6

.000

.034

.069

.103

.138 0

13787 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 100,000 Trials 2,056 Outliers Forecast: Rupture Rate Frequency Chart Fail ures per Susceptible Weld Year Mean = 1.45E-4

.000

.017

.034

.052

.069 0

6899 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 100,000 Trials 2,647 Outliers Forecast: CE RC S TF F AILURES 0.00E+00 1.00E-03 2.00E-03 3.00E-03 4.00E-03 5.00E-03 6.00E-03 7.00E-03 8.00E-03 9.00E-03 1.00E-02 0.00E+0 0

1.00E-012.00E-013.00E-014.00E-015.00E-016.00E-017.00E-018.00E-019.00E-011.00E+0 0

Condition al Probability of Rupture Giv en Failure Prob ability Densit y Prior Posteri or

Frequency Chart Ruptures per Suscepti bl e Weld Year Mean = 7.26E-6

.000

.034

.069

.103

.138 0

13787 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 100,000 Trials 2,056 Outliers Forecast: Rupture Rate Frequency Chart Fail ures per Susceptible Weld Year Frequency Chart Ruptures per Suscepti bl e Weld Year Mean = 7.26E-6

.000

.034

.069

.103

.138 0

13787 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 100,000 Trials 2,056 Outliers Forecast: Rupture Rate Frequency Chart Fail ures per Susceptible Weld Year Mean = 1.45E-4

.000

.017

.034

.052

.069 0

6899 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 100,000 Trials 2,647 Outliers Forecast: CE RC S TF F AILURES Mean = 1.45E-4

.000

.017

.034

.052

.069 0

6899 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 100,000 Trials 2,647 Outliers Forecast: CE RC S TF F AILURES 0.00E+00 1.00E-03 2.00E-03 3.00E-03 4.00E-03 5.00E-03 6.00E-03 7.00E-03 8.00E-03 9.00E-03 1.00E-02 0.00E+0 0

1.00E-012.00E-013.00E-014.00E-015.00E-016.00E-017.00E-018.00E-019.00E-011.00E+0 0

Condition al Probability of Rupture Giv en Failure Prob ability Densit y Prior Posteri or

P(RF) Generic Prior Distribution Beliczy-Schultz Correlation and engineering judgment P(RF) Generic Prior Distribution P(RF) Generic Prior Distribution Beliczy-Schultz Correlation and engineering judgment 12 6/2/11 Pre-Licensing Meeting No.1

Ni-Base Alloys in PWR Primary System Ni Base Alloys in PWR Primary System 13 6/2/11 Pre-Licensing Meeting No.1

Example Results for PWR Hot Leg Pipe T

R V l N l

W ld To Rx Vessel Nozzle Welds 6/2/11 Pre-Licensing Meeting No.1 14

Markov Model Background Markov Model Background Purpose of model is to evaluate the impact of changes to i

i i

f il inspection on pipe failure rates Markov Model originally developed for EPRI RI-ISI Program Applied to 26 plant specific RI-ISI programs in U.S. and South Af i Africa Applied to PBMR to support new ASME Code development for in-service inspections Applied in NUREG 1829 LOCA frequency update Applied in NUREG-1829 LOCA frequency update Currently being applied to address CANDU feeder pipe cracking issue Recently applied to LWRs to guide efforts to reduce internal Recently applied to LWRs to guide efforts to reduce internal flood and HELB contributions to CDF Enhanced version of model developed in DOE/INL RISMC to address aging issues; transition rates based on physics of g g p y failure 15 6/2/11 Pre-Licensing Meeting No.1

MARKOV MODEL OF PIPE ELEMENT MARKOV MODEL OF PIPE ELEMENT S

F

F

Pipe Element States S - success, no detectable damage F - detectable flaw L - detectable leak F

F

R - rupture State Transition Rates L

F

L F

- flaw occurrence rate

- leak failure rate F - rupture failure rate given flaw L - rupture failure rate given leak i

i ISI R

L R

L

- repair rate via ISI exams

- repair rate via leak detection 16 6/2/11 Pre-Licensing Meeting No.1

Example Application of Markov Model to Evaluate St t

i f

Fi P

t ti Pi i Strategies for Fire Protection Piping 1.0E-04 Y-ft.)

1.0E-05 o X (events per ROY Current Study w/ WH Current Study no WH EPRI 1013141 FP NPS > 10" Current Study No WH + Yearly Leak Test Current Study No WH + Quaterly Leak Test 1.0E-06 ater than or Equal to 1 0E 08 1.0E-07 of Rupture Size Grea 1.0E-09 1.0E-08 Frequency o 0.01 0.10 1.00 10.00 100.00 X, Equivalent Break Size (in.)

17 6/2/11 Pre-Licensing Meeting No.1

BWR Recirculation Pipe LOCA Frequency E

l f

NUREG 1860 Example from NUREG-1860 1.0E-04 No ISI/No Leak Inspection No ISI/ Leak Inspection 1/Refueling Outage 1.0E-05 uency/year No ISI/ Leak Inspection 1/Refueling Outage No ISI/ Leak Inspection 1/Week ISI/Leak Inspection 1/Refueling Outage ISI/Leak Inspection 1/Week 1.0E-06 Piping LOCA Frequ 1 0E 08 1.0E-07 WR Recirculation P 1.0E-09 1.0E-08 5

15 25 35 45 55 BW 5

15 25 35 45 55 Plant Age (Years) 18 6/2/11 Pre-Licensing Meeting No.1

Impact of RI-ISI Damage Mechanism Evaluation on RCS Weld Failure Rates Evaluation on RCS Weld Failure Rates (2005 RI-ISI for Koeberg) 6/2/11 Pre-Licensing Meeting No.1 19

Impact of RIM Strategies on SC Susceptible RCS Weld Failure Rate Susceptible RCS Weld Failure Rate (2005 RI-ISI for Koeberg) 6/2/11 Pre-Licensing Meeting No.1 20

Key Interfaces with GSI-191 Evaluation

• Provide estimates of LOCA initiating event Provide estimates of LOCA initiating event frequencies and uncertainties for PRA model (RISKMAN)

- Need to confirm current small, medium, large or revised LOCA size categories P

id l

ti ifi diti l

b bilit

• Provide location specific conditional probability vs. break size information for debris formation/thermal hydraulics model (CASA formation/thermal hydraulics model (CASA GRANDE) 6/2/11 Pre-Licensing Meeting No.1 21

Technical Issues to Address Technical Issues to Address Confirm scope of LOCA events LOCA sensitive piping (e.g. RCS, surge line, ECCS interfaces within p p g (

g g

isolation valves) obviously included Vessel failures and excessive LOCA?

Continue to screen out very small LOCAs?

H t

dd i

t (

CRD l

How to address non-pipe components (e.g. CRD nozzles, manway covers, vessel head seals, RPV instrument lines, head vents)?

How finely to define locations (e.g. specific welds, break orientation?) to resolve debris induced failure variability?

y What is the role of break speed in debris formation?

How to reconcile differences in STP LOCA results vs. NUREG-1829 and NUREG/CR-5750?

What is impact of aging on LOCA frequencies at 40yrs, 60yrs?

22 6/2/11 Pre-Licensing Meeting No.1

Summary of LOCA Frequency Approach

• Utilizes methods and data applied previously in LOCA pp p

y estimation and RI-ISI evaluations

• Capability to specialize frequencies to address key i bl i

ti i

li bilit (

i i

variables impacting pipe reliability (e.g. pipe size, materials, damage mechanisms, inspection status)

• Capability to augment RI-ISI program to optimize NDE Capability to augment RI ISI program to optimize NDE element selection

• Emphasis made on treatment of uncertainties

• Addresses applicable requirements in PRA guides and standards 6/2/11 Pre-Licensing Meeting No.1 23

For more information, please contact:

Karl Fleming fleming@ti-sd.com Bengt Lydell bl @

d bly@scandpower.com Karl Fleming Consulting Service LLC 24 6/2/11 Pre-Licensing Meeting No.1

Back-up Slides Back-up Slides 6/2/11 Pre-Licensing Meeting No.1 25

Estimation of Failure Rates ()

Estimation of Failure Rates ()

• Source data comes from Scandpowers PIPExp comprehensive piping system database database

• Bayes method for uncertainty treatment developed during EPRI RI-ISI program and p

g p

g approved by NRC

• Uncertainties due to scarcity of data, component exposure and modeling component exposure, and modeling accounted for

• Method has been refined and applied over j

t f LWR d PBMR i many projects for LWRs and PBMR since 1997 26 6/2/11 Pre-Licensing Meeting No.1

Bayes Methodology for Failure Rates Bayes Methodology for Failure Rates Number of Leaks And Ruptures for Specific System, Pipe Size, and Damage Mechanism Three Estimates of Component population Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p= 25)

Three Estimates of DM Susceptibility Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p= 25)

Lognormal Distribution Based on engineering Judgment and weld and DM susceptibility Estimates from RISI Number of Leaks And Ruptures for Specific System, Pipe Size, and Damage Mechanism Three Estimates of Component population Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p= 25)

Three Estimates of DM Susceptibility Upper Bound (p=.25)

Best Estimate (p=.50)

Lower Bound (p= 25)

Lognormal Distribution Based on engineering Judgment and weld and DM susceptibility Estimates from RISI 1.20E-02 1.40E-02 1.20E-02 1.40E-02 1.40E-02 1.60E-02 Bayes Update for Three Combinations of Population and DM Susceptibility Lower Bound (p=.25)

Lower Bound (p=.25) 1.20E-02 1.40E-02 1.20E-02 1.40E-02 1.20E-02 1.40E-02 1.40E-02 1.60E-02 1.40E-02 1.60E-02 Bayes Update for Three Combinations of Population and DM Susceptibility Lower Bound (p=.25)

Lower Bound (p=.25) 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probabili ty De nsity Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Rate Probabili ty De nsity Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.20E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probabili ty De nsity Prior Posterior Lower-Lower Update (p=.1)

Best-Best Update (p=.80)

Upper-Upper Update (p=.1)

P(RF), Prob.Rupture Gi F il R

t F

Bayes Posterior Weighting Operation Generic Prior Distribution 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probabili ty De nsity Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Rate 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Rate Probabili ty De nsity Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.20E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probabili ty De nsity Prior Posterior 0.00E+00 2.00E-03 4.00E-03 6.00E-03 8.00E-03 1.00E-02 1.20E-02 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00 Failure Ra te Probabili ty De nsity Prior Posterior Lower-Lower Update (p=.1)

Best-Best Update (p=.80)

Upper-Upper Update (p=.1)

P(RF), Prob.Rupture Gi F il R

t F

Bayes Posterior Weighting Operation Generic Prior Distribution Generic Prior Distribution Frequency Chart Ruptures per Suscepti bl e Weld Year Mean = 7.26E-6

.000

.034

.069

.103

.138 0

13787 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 100,000 Trials 2,056 Outliers Forecast: Rupture Rate Frequency Chart Fail ures per Susceptible Weld Year Mean = 1.45E-4

.000

.017

.034

.052

.069 0

6899 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 100,000 Trials 2,647 Outliers Forecast: CE RC S TF F AILURES 0.00E+00 1.00E-03 2.00E-03 3.00E-03 4.00E-03 5.00E-03 6.00E-03 7.00E-03 8.00E-03 9.00E-03 1.00E-02 0.00E+0 0

1.00E-012.00E-013.00E-014.00E-015.00E-016.00E-017.00E-018.00E-019.00E-011.00E+0 0

Condition al Probability of Rupture Giv en Failure Probability D ensity Prior Posteri or

, Failure Frequency Given Failure

, Rupture Frequency

Frequency Chart Ruptures per Suscepti bl e Weld Year Mean = 7.26E-6

.000

.034

.069

.103

.138 0

13787 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 100,000 Trials 2,056 Outliers Forecast: Rupture Rate Frequency Chart Fail ures per Susceptible Weld Year Frequency Chart Ruptures per Suscepti bl e Weld Year Mean = 7.26E-6

.000

.034

.069

.103

.138 0

13787 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 100,000 Trials 2,056 Outliers Forecast: Rupture Rate Frequency Chart Fail ures per Susceptible Weld Year Mean = 1.45E-4

.000

.017

.034

.052

.069 0

6899 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 100,000 Trials 2,647 Outliers Forecast: CE RC S TF F AILURES Mean = 1.45E-4

.000

.017

.034

.052

.069 0

6899 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 100,000 Trials 2,647 Outliers Forecast: CE RC S TF F AILURES 0.00E+00 1.00E-03 2.00E-03 3.00E-03 4.00E-03 5.00E-03 6.00E-03 7.00E-03 8.00E-03 9.00E-03 1.00E-02 0.00E+0 0

1.00E-012.00E-013.00E-014.00E-015.00E-016.00E-017.00E-018.00E-019.00E-011.00E+0 0

Condition al Probability of Rupture Giv en Failure Probability D ensity Prior Posteri or

, Failure Frequency Given Failure

, Rupture Frequency

P(RF) Generic Prior Distribution Beliczy-Schultz Correlation and engineering judgment P(RF) Generic Prior Distribution P(RF) Generic Prior Distribution Beliczy-Schultz Correlation and engineering judgment 27 6/2/11 Pre-Licensing Meeting No.1

Conditional Probability of Pipe Rupture vs. Break Size 1.0E+00 1.0E-01 1.0E-03 1.0E-02 pture Probability NUREG-1860 BWR Class 1 Fire Protection System Feedwater Condensate 1.0E-05 1.0E-04 Conditional Rup Feedwater Condensate Service Water Circulating Water 1.0E-06 1.0E-07 0.01 0.1 1

10 100 Equivalent Break Size (in.)

28 6/2/11 Pre-Licensing Meeting No.1

MARKOV MODEL OF PIPE ELEMENT S

S

Pipe Element States F

F

S - success, no detectable damage F - detectable flaw L - detectable leak R - rupture L

F L

F State Transition Rates

- flaw occurrence rate

l k f il t

L F

L L

F L

- leak failure rate F - rupture failure rate given flaw L - rupture failure rate given leak

- repair rate via ISI exams

- repair rate via leak detection R

29 6/2/11 Pre-Licensing Meeting No.1

Estimating Input Parameters Estimating Input Parameters

• Degradation related parameters estimated Degradation related parameters estimated from service experience and Bayes models -same as those used for the base models same as those used for the base failure rates

• Test and inspection parameters estimated

• Test and inspection parameters estimated using simple and easy to quantify models 30 6/2/11 Pre-Licensing Meeting No.1

Modeling Impact Of NDE Inspections (ISI) g p

p

(

)

• Capture by the repair rate for flaws P

P F I F D

T T

F I F D I

R where:

- PFI = probability that segment element with flaw will be inspected PFD= probability that flaw is detected given inspection FD p obab y

a a

s de ec ed g e spec o

- TI

= mean time between inspections

- TR = mean time to repair after detection 31 6/2/11 Pre-Licensing Meeting No.1

Modeling Impact Of Leak Tests and Inspections Inspections

• Capture by the repair rate for leaks PLD

T T

LD LI R

(

)

where:

- PLD= probability that leak is detected given inspection

- TI

= mean time between inspections I

ea e be ee spec o s

- TR = mean time to repair after detection 32 6/2/11 Pre-Licensing Meeting No.1

How the Model is Solved How the Model is Solved

• Model used to set up coupled differential equations of t t state

• Closed form solution of equations was performed to enable uncertainty analysis on input parameters

• Hazard rate calculated from state probabilities - yields age dependent rupture frequencies t

dr t

dr

}

{

1

}

{

1

• Hazard rate normalized to create inspection factors in

dt t

dr t

r dt t

dr t

r t

h

}

{

}

{

1 1

}

{

}

{

1

}

{

Hazard rate normalized to create inspection factors in pipe reliability model

• Spreadsheet is used to perform sensitivity analyses to optimize the inspection program optimize the inspection program 33 6/2/11 Pre-Licensing Meeting No.1

State Probabilities for PWR Weld bj Th l F i

subject to Thermal Fatigue 1 0E+00 1.0E-01 1.0E+00 No Damage 1 0E 03 1.0E-02 OBABILITY NO DETECTABLE DAMAGE Probability Close to 1; Sum of all Probabilities

= 1.0 1.0E-04 1.0E-03 STATE PRO DETECTABLE FLAWS DETECTABLE LEAK RUPTURE 1.0 1 0E 06 1.0E-05 1.0E-06 0

10 20 30 40 50 60 70 80 90 100 YEARS INTO PLANT LIFE 34 6/2/11 Pre-Licensing Meeting No.1

Hazard Rate for RCP Weld Subject to Thermal Fatigue 3.5E-07 3.0E-07 2.0E-07 2.5E-07 QUENCY/YEAR h{t}, Time Dependent Hazard Rate 1.0E-07 1.5E-07 RUPTURE FREQ hSS; Steady State Hazard Rate 40-year Average Annual Rupture Probability 5.0E-08 0.0E+00 0

20 40 60 80 100 120 140 160 180 200 TIME IN YEARS 35 6/2/11 Pre-Licensing Meeting No.1

BWR Recirculation Pipe LOCA Frequency from NUREG-1860 1.0E-04 1.0E-05 cy/year No ISI/No Leak Inspection No ISI/ Leak Inspection 1/Refueling Outage No ISI/ Leak Inspection 1/Week ISI/Leak Inspection 1/Refueling Outage ISI/Leak Inspection 1/Week 1.0E-06 ng LOCA Frequenc 1.0E-07 Recirculation Pipin 1.0E-08 BWR R 1.0E-09 5

15 25 35 45 55 Plant Age (Years) 36 6/2/11 Pre-Licensing Meeting No.1

INDEPENDENT REVIEWS OF MARKOV MODEL MARKOV MODEL University of Maryland

- Independent review sponsored by EPRI

- Validated Markov model and Bayes procedure for estimation of Validated Markov model and Bayes procedure for estimation of failure rates

- Found that method was technically sound EdF Benchmark of Markov Model Solution Los Alamos National Laboratory

- Independent review by Martz sponsored by NRC

- Found that method was technically sound Found that method was technically sound USNRC Staff Evaluation found that method was acceptable for use in RISI evaluations under RG 1.178 37 6/2/11 Pre-Licensing Meeting No.1

RI-ISI Results For CDF Change Koeberg UNIT 1 1.700 1.800 1.300 1.400 1.500 1.600 0.900 1.000 1.100 1.200 due to RISI ceptance Criterion)

RISI No Inspection 0.500 0.600 0.700 0.800 Change in CDF action of CDF Acc 0.100 0.200 0.300 0.400 (Fra

-0.200

-0.100 0.000 ARE ASG EAS GCT PTR RCP RCV RIS RPE RRA VVP 38 6/2/11 Pre-Licensing Meeting No.1

Thermohydraulic (T/H) and Downstream Effects Downstream Effects Analyses Overview Risk Informed GSI-191 Resolution Thursday, June 2, 2011 y,

8:00 a.m. - 5:00 p.m.

Public Meeting with STP Nuclear Operating Company Company Texas A&M University 6/2/11 Pre-Licensing Meeting No.1 1

Table of Content Risk Informed GSI-191 Table of Content

Thermohydraulic (T/H) Simulations Purpose Outcome Additional Requirements Software Proposed Approach D

t Eff t M d l (DEM)

Downstream Effect Model (DEM)

Purpose Outcome CFD Simulations CFD Simulations RELAP5-3D Core Nodalization 6/2/11 Pre-Licensing Meeting No.1 2

Risk Informed GSI-191 Thermohydraulic (T/H)

Thermohydraulic (T/H)

Simulations 6/2/11 Pre-Licensing Meeting No.1 3

Purpose Risk Informed GSI-191 Purpose Simulations of the STP Reactor Cooling System (RCS) thermohydraulic response will be performed to support thermohydraulic response will be performed to support the project objectives.

The response of the system under selected accident scenarios will be analyzed, including:

LOCA:

S ll LOCA (

2)

- Small LOCA (< 2)

- Medium LOCA (2 ÷ 6)

- Large LOCA (>6) g

(

)

6/2/11 Pre-Licensing Meeting No.1 4

Outcome Risk Informed GSI-191 Outcome

1. Determine the values of the main thermohydraulic parameters which are expected to affect the debris parameters which are expected to affect the debris generation and transport inside the containment as a function of the time during the selected accidents.

2. Provide the boundary conditions for the Jet Model Development (mass flow

rate, quality, velocity components).

3. Estimate the effects of the debris deposition in the sump screen on the system response during accident requiring long term cooling.

6/2/11 Pre-Licensing Meeting No.1 5

Additional Requirements Risk Informed GSI-191 Additional Requirements Analysis will be performed with:

1) Different Break Locations.

2) Different Break Sizes.

Sensitivity Analysis (SA):

f C di i hi h i

l i i

i d Range of Conditions at which Recirculation is required.

Time to Recirculation.

6/2/11 Pre-Licensing Meeting No.1 6

Software Risk Informed GSI-191 Software RELAP5 will be used to perform the simulations required for the TH section.

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. RELAP5-3D is a successor to the RELAP5/MOD3 code which was developed for the Nuclear Regulatory Commission.

Department of Energy sponsors the code extensions in RELAP5-3D (DOE Office of Fusion Department of Energy sponsors the code extensions in RELAP5 3D (DOE Office of Fusion Energy Sciences, Savannah River Laboratory, Bettis Atomic Power Laboratory, the International RELAP5 Users Group (IRUG), and the Laboratory Directed Research and Development Program at the INL[1]).

The code provides fully integrated, coarse-mesh multi-dimensional thermal-hydraulic and kinetic modeling capability.

It includes a multi-dimensional component to approximately model the multi-dimensional flow behavior that can be exhibited in regions of a LWR system such as lower plenum flow behavior that can be exhibited in regions of a LWR system such as lower plenum, core, upper plenum and downcomer [1].

[1] RELAP5-3D User Manual Volume: 1 6/2/11 Pre-Licensing Meeting No.1 7

[1] RELAP5-3D User Manual Volume: 1.

Software Risk Informed GSI-191 Software DAKOTA will be coupled with RELAP5 to perform the Sensitivity Analysis and Uncertainty Quantification of the thermohydraulic calculations.

DAKOTA is a tool with wide usage at the DOE NNSA labs and the DAKOTA is a tool with wide usage at the DOE-NNSA labs and the wider community.

It has a strong foundation of verification and validation.

g n1 Dakota RELAP5 n2 RELAP5 Output nx Output +

6/2/11 Pre-Licensing Meeting No.1 8

Input Variables Output +

Uncertainty

Proposed Approach 1/2 Risk Informed GSI-191 Proposed Approach 1/2 STEP 1: RELAP5 Steady-State Input Certification 1A) I t P ti 1A) Input Preparation 1B) Input Certification 1C) Documentation Preparation 1D) St d

St t lt l

i 1D) Steady-State results analysis STEP 2: LOCA Input Certification 2A) Input Preparation 2B) Input Certification (Certified STP RETRAN input and MAAP documents used) 2C) Documentation Preparation 2D) Reference LOCA results analysis (comparison with available RELAP5-PWR simulations of LOCA) 6/2/11 Pre-Licensing Meeting No.1 9

Proposed Approach 2/2 Risk Informed GSI-191 Proposed Approach 2/2 STEP 3: 3D Core Nodalization Certification STEP 3: 3D Core Nodalization Certification 3A) 3D Core Nodalization Input Preparation 3B) LOCA analysis and comparison with 1D results STEP 4: Analysis of the Required Cases 6/2/11 Pre-Licensing Meeting No.1 10

STEP 1: RELAP5 Steady State Input Certification Risk Informed GSI-191 STEP 1: RELAP5 Steady-State Input Certification RELAP5 nodalization was not available for the STP Plant.

The plant nodalization was prepared using the existing RETRAN input file (previously certified), and the MAAP documents provided by STP.

The typical PWR Westinghouse nodalization proposed in the RELAP5 users manual was also considered [2].

The input will be described in a dedicated documentation showing the main input features and the references used.

Simulation results are compared with the provided steady-state conditions in order to validate the certification efforts.

2 A

3 l

l 6/2/11 Pre-Licensing Meeting No.1 11

[2] RELAP5-3D User Manual Volume: 5.

STEP 2: LOCA Input Certification Risk Informed GSI-191 STEP 2: LOCA Input Certification The input will be prepared starting from the certified steady-state input.

S (A

S S) i SI components (Accumulators, HPIS, LPIS) and the system control are being implemented in the input.

Certified RETRAN input file, MAAP documentation and SFAR are used in the LOCA p

input preparation.

The input will be certified using existing simulations of LOCA (e.g., small, medium d l b

k i

) [

3]

and large break sizes) [e.g., 3].

[3] NUREG/CR-6770 LA-UR-01-5561. GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System and Containments to Selected Accident Sequences 6/2/11 Pre-Licensing Meeting No.1 12 q

STEP 3: 3D Core Nodalization Certification Risk Informed GSI-191 STEP 3: 3D Core Nodalization Certification 3D Core nodalization is required for Downstream Effects Analysis 1 D Core 6/2/11 Pre-Licensing Meeting No.1 13 1-D Core

STEP 3: 3D Core Nodalization Certification Risk Informed GSI-191 STEP 3: 3D Core Nodalization Certification 3D Core nodalization is required for Downstream Effects Analysis 3 D Core 6/2/11 Pre-Licensing Meeting No.1 14 3-D Core

STEP 4: Analysis of the Cases Risk Informed GSI-191 STEP 4: Analysis of the Cases Break Locations:

Break Locations:

Selected to account of various break locations, etc. to study the system response.

Break Sizes:

Break Sizes:

Small, Medium and Large LOCA will be analyzed.

Sensitivity analysis will determine the minimum break size at which recirculation is required.

q 6/2/11 Pre-Licensing Meeting No.1 15

Risk Informed GSI-191 Downstream Effects Model Downstream Effects Model (DEM) Development 6/2/11 Pre-Licensing Meeting No.1 16

Purpose Risk Informed GSI-191 Purpose Debris may be small enough to pass through the sump screen and reach the reactor core screen and reach the reactor core.

• The debris transport and deposition through The debris transport and deposition through downcomer, lower plenum and reactor core will be investigated.

• The effects of the debris deposition on the thermohydraulic response of the system (core & fuel thermohydraulic response of the system (core & fuel temperature distribution) will be studied.

6/2/11 Pre-Licensing Meeting No.1 17

Outcome Risk Informed GSI-191 Outcome

1. Debris transport and deposition model in selected locations in the reactor vessel (downcomer lower locations in the reactor vessel (downcomer, lower plenum, fuel assemblies).

2 Simulations of the thermal hydraulics system

2. Simulations of the thermal hydraulics system response under conditions identified in 1 (response parameters such as fuel temperature distribution).

p p

)

6/2/11 Pre-Licensing Meeting No.1 18

Software Risk Informed GSI-191 Software

Computational Fluid Dynamics (CDF) codes will be used to predict the

Computational Fluid Dynamics (CDF) codes will be used to predict the debris transport and deposition in the core

RELAP5-3D system code will be used to perform the thermal hydraulics simulations of the reactor system. The core will be modeled with multidimensional components to describe the flow behavior within core.

6/2/11 Pre-Licensing Meeting No.1 19

CFD Simulations Risk Informed GSI-191 CFD Simulations Approach

• Single Phase Flow Analysis

• Isothermal Conditions

• Discrete Phase Method (DPM) h t d l th d b i approach to model the debris inside the reactor

• 1-way coupling 6/2/11 Pre-Licensing Meeting No.1 20 Ref: Timothy D. Sande, Alionscience

RELAP 3D C N d li i

Risk Informed GSI-191 RELAP5-3D Core Nodalization Loop 1 Loop 2 Loop 4 Loop 3 6/2/11 Pre-Licensing Meeting No.1 21

List of References

[O] Document Available; [X] Document not yet Available

[ ]

[ ]

y

[O] RELAP5-3D Code Manuals, Idaho National Laboratory, INEEL-EXT-98-00834, Revision 2.4, June 2005.

[O] RELAP5/MOD3.3 Code Manual Volume III: Developmental Assessment Problems, NUREC/CR-5535/Rev 1-Vol III, December 2001.

5535/Rev 1 Vol III, December 2001.

[O] NUREG/CR-6770 LA-UR-01-5561. GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System and Containments to Selected Accident Sequences.

[O] Moore RL, Sloan SM, Schultz RR, Wilson GE, RELAP5/MOD3 Code Manual - Summary and Reviews of Independent Code Assessment Reports, NUREG/CR-5535, INEL-95/0174 Vol. 7, Rev. 1, Idaho National Engineering Laboratory, October 1996.

Idaho National Engineering Laboratory, October 1996.

[O] NUREG/IA-0242, Qualification of the Three-Dimensional Thermal Hydraulic Model of TRACE using Plant Data

[O] Aksan N, Selected Examples of Natural Circulation for Small Break LOCA and Some Severe Accidents (Presentation), IAEA Course on Natural Circulation in Water-Cooled Nuclear Power Plants, International Centre for Theoretical Physics (ICTP) Trieste Italy June 25-29 2007 International Centre for Theoretical Physics (ICTP), Trieste, Italy, June 25 29, 2007.

[O] Aksan N, International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA),

Science and Technology of Nuclear Installations Vol. 2008, ID: 814572, pp20, 2008.

[O] Bae B-U, Lee KH, Kim YS, Yun BJ, Park GC, Scaling Methodology for a Reduced-Height Reduced-Pressure Integral Test Facility to Investigate Direct Injection Line Break SBLOCA, Nuclear Engineering and Design Vol 238 pp 2197-2205 2008 Engineering and Design Vol. 238, pp 2197 2205, 2008.

[O] Bayless PD et al., Severe Accident Natural Circulation Studies at INEL, NUREG/CR-6285, Idaho National Engineering Laboratory.

[O] Boyack BE, Lime JF, Intermediate-Break LOCA Analyses for the AP600 Design, LA-UR-95-1785, Los Alamos National Laboratory, 1995.

[X] Brittain I Aksan SN OECD-LOFT Large Break LOCA Experiments: Phenomenology and Computer 6/2/11 Pre-Licensing Meeting No.1 22

[X] Brittain I, Aksan SN, OECD-LOFT Large Break LOCA Experiments: Phenomenology and Computer Code Analyses, AEEW-TRS-1003, PSI-Bericht Nr. 72, Paul Scherrer Institute, August 1990

List of References

[O] Burtt JD, Crowton SA, International Standard Problem 13 (LOFT Experiment 2-5) (Preliminary Comparison Report), CSNI Report 101, Idaho National Engineering Laboratory, April 1983.

[O] Chung BD, Lee YJ, Hwang TS, Lee WJ, Lee SY, Quantification of Reactor Safety Margins for Large Break LOCA with Application of Realistic Evaluation Methodology, Journal of Korean Nuclear Society, Vol. 26, 1994.

[O] Clement P Chataing T Deruaz R OECD/NEA/CSNI International Standard Problem No 27-BETHSY

[O] Clement P, Chataing T, Deruaz R, OECD/NEA/CSNI International Standard Problem No. 27 BETHSY Experiment 9.1B - 2 Cold Leg Break without HPSI and with Delayed Ultimate Procedure (Comparison Report) Volume 1, NEA/CSNI/R(92)20, November 1992.

[O] Clement P, Chataing T, Deruaz R, OECD/NEA/CSNI International Standard Problem No. 27-BETHSY Experiment 9.1B - 2 Cold Leg Break without HPSI and with Delayed Ultimate Procedure (Comparison Report) Volume 2 NEA/CSNI/R(92)20 Committee on the Safety of Nuclear Installation (CSNI) OECD Nuclear Report) Volume 2, NEA/CSNI/R(92)20, Committee on the Safety of Nuclear Installation (CSNI) OECD Nuclear Energy Agency (NEA), November 1992.

[O] Davis CB, Assessment of the RELAP5 Multi-Dimensional Component Model Using Data from LOFT Test L2-5, INEEL-EXT-97-01325, Idaho National Engineering and Environmental Laboratory, January 1998.

[O] de Crécy A, P. Bazin P, BEMUSE Phase III Report: Uncertainty and Sensitivity Analysis of the LOFT L2-5 Test NEA/CSNI/R(2007)4 CSNI OECD NEA October 2007 Test, NEA/CSNI/R(2007)4, CSNI OECD NEA, October 2007.

[X] Fletcher CD, Kullberg CM, Break Spectrum Analysis for Small Break Loss-of-Coolant Accidents in a ESAR-3S Plant, NUREG/CR-4384, EGG-2416, Idaho National Engineering Laboratory, September 1985.

[O] Clement P, Chataing T, Deruaz R, OECD/NEA/CSNI International Standard Problem No. 27-BETHSY Experiment 9.1B - 2 Cold Leg Break without HPSI and with Delayed Ultimate Procedure (Comparison Report) Volume 1 NEA/CSNI/R(92)20 November 1992 Report) Volume 1, NEA/CSNI/R(92)20, November 1992.

[O] Clement P, Chataing T, Deruaz R, OECD/NEA/CSNI International Standard Problem No. 27-BETHSY Experiment 9.1B - 2 Cold Leg Break without HPSI and with Delayed Ultimate Procedure (Comparison Report) Volume 2, NEA/CSNI/R(92)20, Committee on the Safety of Nuclear Installation (CSNI) OECD Nuclear Energy Agency (NEA), November 1992.

6/2/11 Pre-Licensing Meeting No.1 23

List of References

[O] D i

CB A t

f th RELAP5 M lti Di i

l C t M d l U i D t f

LOFT T t

[O] Davis CB, Assessment of the RELAP5 Multi-Dimensional Component Model Using Data from LOFT Test L2-5, INEEL-EXT-97-01325, Idaho National Engineering and Environmental Laboratory, January 1998.

[O] de Crécy A, P. Bazin P, BEMUSE Phase III Report: Uncertainty and Sensitivity Analysis of the LOFT L2-5 Test, NEA/CSNI/R(2007)4, CSNI OECD NEA, October 2007.

[X] Fletcher CD, Kullberg CM, Break Spectrum Analysis for Small Break Loss-of-Coolant Accidents in a ESAR 3S Pl t NUREG/CR 4384 EGG 2416 Id h N ti l E i

i L b t

S t

b 1985 ESAR-3S Plant, NUREG/CR-4384, EGG-2416, Idaho National Engineering Laboratory, September 1985.

[O] Griggs DP, Liebmann ML, Comparison of TRAC and RELAP5 Reactor System Calculations for a DEGB LOCA in K-14.1. WSRC-TR-90-393, Westinghouse Savanah River Laboratory, September 1990.

[O] Jeong JJ, Sim SK, Ban CH, Park CE, Assessment of the COBRA/RELAP5 Code Using the LOFT L2-3 Large-Break Loss-of-Coolant Experiment, Annals of Nuclear Energy Vol. 14, pp 1171-1182, 1997.

[O] Kim Y-S, Bae B-U, Park G-C, Integral Loop Test and Assessment of Modified RELAP5/MOD3.3 for RCS Coolant Inventory during LBLOCA, Nuclear Engineering and Design Vol. 237, pp182-198, 2007.

[O] Kukita Y et al., OECD/NEA/CSNI International Standard Problem No. 26-ROSA-IV Cold-Leg Small Break LOCA Experiment (Comparison Report) Volume 2, NEA/CSNI/R(91)13, CSNI OECD NEA, February 1992.

[O] Liang TKS, Chang C-J, Hung H-J, Development of LOCA Licensing Calculation Capability with RELAP5-3D in accordance with Appendix K of 10 CFR 50.46. Nuclear Engineering and Design Vol. 211, pp69-84, 2002.

[O] OECD, Lessons Learned from OECD/CSNI ISP on Small Break LOCA, OCDE/GD(97)10, OECD, July 1996.

[O] Petruzzi A, DAuria F, BEMUSE Phase II Report: Re-Analysis of the ISP-13 Exercise, Post Test Analysis of the LOFT L2-5 Test Calculation, NEA/CSNI/R(2006)2, CSNI OECD NEA, November 2005.

[O] Reventos F, Perez M, Batet L, Pericas R, BEMUSE Phase IV Report: Simulation of a LB-LOCA in ZION Nuclear Power Plant (Volume 1), NEA/CSNI/R(2008)6, CSNI OECD NEA, November 2008.

6/2/11 Pre-Licensing Meeting No.1 24

List of References

[O] Reventos F, Perez M, Batet L, Pericas R, BEMUSE Phase IV Report: Simulation of a LB-LOCA in ZION Nuclear Power Plant (Volume 2), NEA/CSNI/R(2008)6, CSNI OECD NEA, November 2008.

[O] Reventos F, Perez M, Batet L, Pericas R, BEMUSE Phase IV Report: Simulation of a LB-LOCA in ZION Nuclear Power Plant (Volume 3), NEA/CSNI/R(2008)6, CSNI OECD NEA, November 2008.

[O] Suh JK Bang YS Kim HJ Assessment of RELAP5/MOD3 2 2gamma with the LOFT L9-3 Experiment

[O] Suh JK, Bang YS, Kim HJ, Assessment of RELAP5/MOD3.2.2gamma with the LOFT L9 3 Experiment Simulating an Anticipated Transient Without Scram, Korea Institute of Nuclear Energy, November 2000.

[O] Takeda T, Asaka H, Nakamura H, Analysis of the OECD/NEA ROSA Project Experiment Simulating a PWR Small Break LOCA with high-power natural circulation. Annals of Nuclear Energy Vol. 36, pp 386-392, 2009.

[O] Takeda T, Asaka H, Suzuki M, Nakamura H, RELAP5 Analysis of ROSA/LSTF Vessel Upper Head Break LOCA Experiment The 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-LOCA Experiment, The 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12), Pittsburgh, Pennsylvania, U.S.A. September 30-October 4, 2007.

General

[O] IAEA, Safety Report Series No. 52 - Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation International Atomic Energy Agency 2008 Evaluation, International Atomic Energy Agency, 2008.

[O] IAEA, Safety Standards Series No. NS-G-1.2 - Safety Assessment and Verification for Nuclear Power Plants, International Atomic Energy Agency, 2001.

[O] IAEA, Safety Report Series No. 30 - Accident Analysis for Nuclear Power Plants with Pressurized Water Reactors, International Atomic Energy Agency, 2003.

6/2/11 Pre-Licensing Meeting No.1 25