ML11157A010
| 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
RiskInformed GSI191 Closure Plan Risk Informed GSI191 Resolution Thursday, June 2, 2011 8:00 a.m. 5:00 p.m.p Public Meeting with STP Nuclear Operating Company Rick Grantom South Texas Project 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
6/2/11 Pre-Licensing Meeting No.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 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, 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 off R Region Regulatory l G Guide id 11.174) 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 rupture t off th the llargestt pipei iin theth reactor t coolant l t system t (RCS)
(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 risk-informed informed model as an acceptable alternate evaluation model to calculate debris generation, 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 exemptionti and d th the ttechnical h i l methods th d used d iin th the calculations.
l l ti It would ld conform f with ith 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
Risk Informed GSI-191 Section View 6/2/11 Pre-Licensing Meeting No.1 2
Risk Informed GSI-191 Operating Deck 6/2/11 Pre-Licensing Meeting No.1 3
Risk Informed GSI-191 SG Compartment - View 1 6/2/11 Pre-Licensing Meeting No.1 4
Risk Informed GSI-191 SG Compartment - View 2 6/2/11 Pre-Licensing Meeting No.1 5
Risk Informed GSI-191 SG Compartment Floor 6/2/11 Pre-Licensing Meeting No.1 6
Risk Informed GSI-191 Pool Elevation - View 1 6/2/11 Pre-Licensing Meeting No.1 7
Risk Informed GSI-191 Pool Elevation - View 2 6/2/11 Pre-Licensing Meeting No.1 8
Risk Informed GSI-191 Pool Elevation - View 3 6/2/11 Pre-Licensing Meeting No.1 9
Risk Informed GSI-191 Piping and Insulation 6/2/11 Pre-Licensing Meeting No.1 10
Risk Informed GSI-191 RCS System Piping and Insulation 6/2/11 Pre-Licensing Meeting No.1 11
Risk Informed GSI-191 Insulation on Hangers and Valves 6/2/11 Pre-Licensing Meeting No.1 12
Risk Informed GSI-191 Work Points for Hangers/Valves/Welds 6/2/11 Pre-Licensing Meeting No.1 13
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 Risk Informed GSI-191 Resolution Thursday, June 2, 2011 8:00 a.m. - 5:00 p.m.
Public Meeting with STP Nuclear Operating Company Presented by David H. Johnson, Sc.D.
6/2/11 Pre-Licensing Meeting 1 S:\SharedFiles\Local\Pubs\VUGRAFS\DHJ\STP June 2, 2011\PRA Modeling - STP 6-2-11 No.1
Risk Informed GSI-191 Discussion of Status and Direction of PRA Modeling 6/2/11 Pre-Licensing Meeting 2 No.1
Risk Informed GSI-191 Near Term PRA Activities
- Develop p Expanded p LOCA Event Trees
- Detailed Sequence of Events Associated with Potential Sump Performance Phenomena
- Develop Necessary Logic Structures to Represent Results from Other Team Members
- Support Uncertainty Calculation
- Integration of New Analyses
- Utilize Analyses from Other Team Members
- Use PRA Framework to Assist in Technical Direction of Specialized Analyses
- Use Existing PRA Model to Identify Potential Analysis y Boundaries 3
6/2/11 Pre-Licensing Meeting No.1
Risk Informed GSI-191 LLOCA Event Sequence Diagram 4
Risk Informed GSI-191 LLOCA Event Tree 5
Risk Informed GSI-191 SI Recirculation Fault Tree (Top) 6
Risk Informed GSI-191 Current and Future PRA Models
- Existing PRA
- Does Not Explicitly Address Sump Plugging Phenomena
- Does D IInclude l d AB Basic i E Event (1E(1E-5)
- 5) to Represent R Unavailability U il bili off Sump; Required for All Sequences Involving Recirculation
- At-Power Importance:
- 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
- Assist A i t in i D Documentation t ti off EffEffortt
- Eventual Incorporation into PRA Model of Record
- Model May Be Higher Level, As Appropriate
- Could Result in Revised Initiator Groupings 7
6/2/11 Pre-Licensing Meeting No.1
Risk Informed GSI-191 Sensitivityy Analyses y
8
Risk Informed GSI-191 Logic Changes Underway
- Break Out Sump Blockage Logic
- Add Logic to Represent In-Core Downstream Phenomena Downstream
- Add Logic to Represent Potential O
Operational ti l St Strategies t i
Risk Informed GSI-191 Break out Sump Blockage Logic
- 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 Fail
- Based on Debris Amount and Flow
- System Operation State 6/2/11 Pre-Licensing Meeting 10 No.1
Risk Informed GSI-191 SOUTH TEXAS PROJECT LARGE BREAK LOCA EVENT SEQUENCE DIAGRAM CURRENT MODEL LARGE REACTOR POWER LOCA DECREASED CORE BECOMES SUBCRITICAL DUE TO VOIDING. RCS IA IB IC IA IB IC PRESSURE BELOW LHSI SHUTOFF HEAD CHI SI RA RB RC RX SSPS PZP ACP DCP CCW SI38A SI38B SI38C AI PA PB PZ 4 6 10 11 12 13 15 H LEG 1 2 3 INJECTION COMMON ACCUMULATOR RWST/ECCS CORE HOT LEG STABLE TO LHSI AND SAFETY ESFAS LHSI REMAINS RHR HX ACCUMULATOR INJECTION COMMON RECIRCULATION CONDITION SUBCRITICAL DUE TO INJECTING BORATED WATER FROM RWST OR ACP DCP LOSS OF SI PUMP 5 HI NPSH MANUAL SI 7
OR MANUAL START HHSI M CCW ACP DCP 8
14 CONTAINMENT SPRAY CF RCFC WET CS LT2 EARLY DRY LOW PRESSURE LT2 LATE CS CH3 WET LOW PRESSURE ACP DCP SS 9 8 SSPS, Q
SEQUENCER CONTAINMENT COMMON SPRAY WET LT2 EARLY DRY LOW PRESSURE 11
Risk Informed GSI-191 12
Risk Informed GSI-191 Add Logic to Represent Downstream Effect
- Current Model
- Fuel Damage Due to Material Passing Screens Not Included
- Desired Model
- Include Scenarios Addressing In In-Core Core Phenomena Due to Material Passing Strainer
- Will Assign Any Damage Scenarios to Unique Damage States 6/2/11 Pre-Licensing Meeting 13 No.1
Risk Informed GSI-191 14
Risk Informed GSI-191 Add Logic g to Represent p Potential Operational Strategies
- Current Model
- Follows Current Procedures
- Desired Model
- Explicitly Represent Actions That Could Influence Amount of Debris Arriving at Individual Screens (e.g., Cycling RHR Train) 6/2/11 Pre-Licensing Meeting 15 No.1
Risk Informed GSI-191 16
Risk Informed GSI-191 Differentiate LOCA Initiators Based on Potential for Sump Blockage
- Current Model
- Single Initiator Categories for Small (Isolatable and Nonisolatable), Medium and Large g LOCAs
- Desired Model
- Identify Explicitly Those Specific Break Locations That Impact Debris Phenomena 6/2/11 Pre-Licensing Meeting 17 No.1
Risk Informed GSI-191 18
Risk Informed GSI-191 Incorporating Results from Detailed A l Analyses
- As results of detailed analyses y become available, their insights will be reflected in PRA model
- Interpretation p of results of detailed analyses y to support pp scenario-specific basic events and event tree branching likelihoods
- Potential changes to model structure possible
- Support pp scenario-specific p basic event and event 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 g 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
- Issues and strategies for resolution 6/2/11 Pre-Licensing Meeting No.1 2
Risk Informed GSI-191 LOCA Frequency Project Team
- K l Fl Karl Fleming, i P President id t 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 j on RI-ISI, IFPRA, and EPRI pipe failure data
- Bengt Lydell, VP, Sr. Principal Consultant, Scandpower Inc.
- Passive component reliability, IFPRA, RI-ISI, OECD/NEA Consultant
- NUREG-1829 Team Member
- Principal P i i l author h off EPRI projects j on RI RI-ISI, ISI IFPRA, IFPRA andd EPRI pipe i failure f il data d
- Fred Simonen, Sr. Principal Consultant, Scandpower Inc.
- ASME Fellow
- Fracture mechanics, mechanics structural integrity, integrity PTS, PTS RIRI-ISI 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 6/2/11 Pre-Licensing Meeting No.1 3
Risk Informed GSI-191 Team Relevant Experience
- 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 g risk impact p evaluation for STP RI-ISI p pilot study y
- 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 6/2/11 Pre-Licensing Meeting No.1 4
Risk Informed GSI-191 LOCA Frequencies Team S
Scope off Work W k
- Incorporate insights from previous work on LOCA frequencies
- Characterize LOCA initiating events and there frequencies with respect to:
- Specific components, materials, dimensions
- Specific locations
- Range g of break sizes
- 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
- Break B k characterization h t i ti ffor evaluation l ti off debris d b i fformation ti
- Participate in NRC workshops 5 6/2/11 Pre-Licensing Meeting No.1
LOCA Frequency Technical A
Approach h
- 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 Utilize Markov k model d l tto evaluate l t iimpactt off NDE element l t selection l ti ffrom RI RI-ISI 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 p manner
- 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 6 6/2/11 Pre-Licensing Meeting No.1
Key Results from NUREG-1829 NUREG 1829 7 6/2/11 Pre-Licensing Meeting No.1
LOCA IE Frequency Model 1 of 2 F ( LOCAx ) mi ix (1) i jx ix ik P( Rx Fik ) I ik (2) k Where:
F ( LOCAx ) Frequency of LOCA of size x, per reactor calendaryear; subject to epistemic i t i uncertainty t i t calculated l l t d via i M Monte t CCarlo l
mi Number of pipe welds of type i; each type determined by pipe size, weld type, applicable damage mechanisms, and inspection status (leak test and NDE); no uncertainty ix Frequency of rupture of pipe location j belonging to component type i with break size x, subject to epistemic uncertainty calculated via Monte Carlo ik Failure rate per weldyear for pipe component type i due to failure mechanism k; subject to epistemic uncertainty determined by RIISI Bayes y method and Eq. q ((3))
P( Rx Fik ) Conditional probability of rupture of size x given failure of pipe component type i due to damage mechanism k; subject to epistemic uncertainty determined via expert elicitation (NUREG1829)
I ik Integrity management factor for weld type i and failure mechanism k; subject to epistemic uncertainty determined by Monte Carlo and Markov Model 8 6/2/11 Pre-Licensing Meeting No.1
LOCA IE Frequency q y Model 2 of 2 nik nik ik (3) ik f ik N iTi nik Number of failures in pipe component (i.e. weld) type i due to failure mechanism k, very little epistemic uncertainty ik C Component t exposure population l ti for f weldsld off type t i susceptible tibl tto failure mechanism k, subject to epistemic uncertainty determined by expert opinion f ik Estimate of the fraction of the component exposure population for weld type i that is susceptible to failure mechanism kk, subject to epistemic uncertainty; estimated from results of RIISI for population of plants and expert opinion Ni Estimate of the average number of pipe welds of type i per reactor in the applicable reactor years exposure for the data collection; subject to epistemic uncertainty; estimated from results of RIISI for population of plants and expert opinion Ti Total number of reactor years exposure for the data collection for component type i; little or no uncertainty 9 6/2/11 Pre-Licensing Meeting No.1
Estimation of Failure Rates ()
- Source data comes from Scandpowers p PIPExp p
comprehensive piping system database
- Bayes method for uncertainty treatment developed during g EPRI RI-ISI p program g and approved pp by y NRC
- Uncertainties due to scarcity of data, component exposure, and modeling accounted for
- Methods and database recently applied to develop EPRI ppipe p failure data handbook for IFPRA and HELB PRA (EPRI 1021086)
- Methods being refined to address aging as part of DOE/INL Risk-Informed Safetyy Margin g Characterization Project 10 6/2/11 Pre-Licensing Meeting No.1
Volume of Available Data Records f PWR Class for Cl 1 and d 2 Pi Piping i
6/2/11 Pre-Licensing Meeting No.1 11
Bayes Methodology for Failure Rates Bayes Number of Leaks Three Estimates of Three Estimates of And Ruptures for Component population DM Susceptibility Specific System, Lognormal Distribution Pipe Size, and Upper Bound (p=.25) Upper Bound (p=.25) Based on engineering Damage Mechanism Judgment and weld Best Estimate (p=.50) Best Estimate (p=.50) and DM susceptibility Estimates from RISI Lower Bound (p=.25) Lower Bound (p=.25)
Bayes Update for Three Combinations of Population and DM Susceptibility 1.40E -02 1.40E-02 1.60E-02 1.40E-02 1.20E -02 1.20E-02 1.20E-02 1.00E -02 1.00E-02 1.00E-02 8.00E -03 8.00E-03 P r oba abili ty De nsity P r oba abili ty De nsity P r obab bili ty De nsity 8.00E-03 6 00E -03 6.00E 03 6 00E-03 6.00E 03 6.00E-03 4.00E -03 4.00E-03 4.00E-03 Generic Prior 2.00E -03 2.00E-03 2.00E-03 0.00E +00 0.00E+00 0.00E+00 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 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 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 Failur e Ra te Prior P osterior Failure Ra te Prior P osterior P rior P osterior Distribution Failur e Ra te Lower-Lower Update (p=.1) Best-Best Update (p=.80) Upper-Upper Update (p=.1)
Bayes Posterior Weighting Operation P(RF), Prob.Rupture
, Failure Frequency Given Failure , Rupture Frequency Forecast C E RC S TF F AILU RES Forecast: Forecast: Rupture Rate 1.00E-02 100,000 Trial s Frequ en cy Ch art 2,647 O utliers 9.00E-03 100,000 Trials Frequency Chart 2,056 Outliers
.069 6899
.138 13787 8.00E-03 Prob abilit y D en sit y 7.00E-03
.052 .103 6.00E-03 5.00E-03
.034 Prior .069 4.00E-03 Posteri or 3.00E-03
.034
.017 2.00E-03 1.00E-03 Mean = 7.26E-6 Mean = 1. 45E-4 .000 0 0.00E+00
.000 0 0.00E+01.00E-01 2.00E-013.00E-01 4.00E-015.00E-016.00E-017.00E-01 8.00E-01 9.00E-011.00E+0 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 0 0 Fail ures per S usc eptible Wel d Year Condition al Prob abilit y of Rupture Giv en F ailur e Ruptur es per Susc epti bl e Weld Year Beliczy-Schultz Correlation and engineering P(RF) Generic Prior judgment 12 6/2/11 Pre-Licensing Meeting No.1 Distribution
Ni-Base Ni Base Alloys in PWR Primary System 13 6/2/11 Pre-Licensing Meeting No.1
Example Results for PWR Hot Leg Pipe T Rx To R Vessel V lNNozzle l W Welds ld 6/2/11 Pre-Licensing Meeting No.1 14
Markov Model Background
- Purpose of model is to evaluate the impact of changes to i
inspection i on pipe i ffailure il 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 NUREG 1829 LOCA frequency update
- Currently being applied to address CANDU feeder pipe cracking issue
- Enhanced version of model developed in DOE/INL RISMC to g g issues;; transition rates based on p address aging physics y of failure 15 6/2/11 Pre-Licensing Meeting No.1
MARKOV MODEL OF PIPE ELEMENT S
Pipe Element States S - success, no detectable damage F - detectable flaw L - detectable leak F R - rupture State Transition Rates
- flaw occurrence rate F L - leak failure rate F - rupture failure rate given flaw L - rupture failure rate given leak L - repairi rate via i ISI exams
- repair rate via leak detection R
6/2/11 Pre-Licensing Meeting 16 No.1
Example Application of Markov Model to Evaluate St t i for Strategies f Fire Fi Protection P t ti Piping Pi i 1.0E-04 Frequency o of Rupture Size Grea ater than or Equal to o X (events per ROY Y-ft.)
Current Study w/ WH 1.0E-05 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 1.0E-07 1 0E 08 1.0E-08 1.0E-09 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
Examplel ffrom NUREG-1860 NUREG 1860 1.0E-04 No ISI/No Leak Inspection No ISI/ Leak Inspection 1/Refueling Outage No ISI/ Leak Inspection 1/Week ISI/Leak Inspection 1/Refueling Outage BW WR Recirculation P Piping LOCA Frequ uency/year 1.0E-05 ISI/Leak Inspection 1/Week 1.0E-06 1.0E-07 1 0E 08 1.0E-08 1.0E-09 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 (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 (2005 RI-ISI for Koeberg) 6/2/11 Pre-Licensing Meeting No.1 20
Key Interfaces with GSI-191 Evaluation
- Need to confirm current small, medium, large or revised LOCA size categories
- P Provide id llocation ti specific ifi conditional diti l probability b bilit vs. break size information for debris formation/thermal hydraulics model (CASA GRANDE) 6/2/11 Pre-Licensing Meeting No.1 21
Technical Issues to Address
- Confirm scope of LOCA events
- LOCA sensitive p piping p g ((e.g.
g RCS,, surgeg line,, ECCS interfaces within isolation valves) obviously included
- Vessel failures and excessive LOCA?
- Continue to screen out very small LOCAs?
- How H tto address dd non-pipe i components t (e.g.
( CRD nozzles, l 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 pp p previously y in LOCA estimation and RI-ISI evaluations
- Capability to specialize frequencies to address key variables i bl iimpacting ti pipei reliability li bilit ((e.g. pipe i size, i
materials, damage mechanisms, inspection status)
- 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 g
fleming@ti-sd.com Bengt Lydell bl @
bly@scandpower.com d
Karl Fleming Consulting Service LLC 24 6/2/11 Pre-Licensing Meeting No.1
Back-up Slides 6/2/11 Pre-Licensing Meeting No.1 25
Estimation of Failure Rates ()
- Source data comes from Scandpowers PIPExp comprehensive piping system database
- Bayes method for uncertainty treatment developed p during g EPRI RI-ISI p program g and approved by NRC
- Uncertainties due to scarcity of data, component exposure exposure, and modeling accounted for
- Method has been refined and applied over many projects j t ffor LWRsLWR and d PBMR since i
26 1997 6/2/11 Pre-Licensing Meeting No.1
Bayes Methodology for Failure Rates Bayes Number of Leaks Three Estimates of Three Estimates of And Ruptures for Component population DM Susceptibility Specific System, Lognormal Distribution Pipe Size, and Upper Bound (p=.25) Upper Bound (p=.25) Based on engineering Damage Mechanism Judgment and weld Best Estimate (p=.50) Best Estimate (p=.50) and DM susceptibility Estimates from RISI Lower Bound (p=
(p=.25)
- 25) Lower Bound (p=
(p=.25) 25)
Bayes Update for Three Combinations of Population and DM Susceptibility 1.40E -02 1.40E-02 1.60E-02 1.40E-02 1.20E -02 1.20E-02 1.20E-02 1.00E -02 1.00E-02 1.00E-02 8.00E -03 8.00E-03 Pr obabili ty De nsity Pr obabili ty De nsity Pr obabili ty De nsity 8.00E-03 6.00E -03 6.00E-03 6.00E-03 4.00E -03 4.00E-03 4.00E-03 Generic Prior 2.00E -03 2.00E-03 2.00E-03 0.00E +00 0.00E+00 0.00E+00 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 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 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 Failur e Ra te Prior P osterior Failure Ra te Prior P osterior P rior P osterior Distribution Failur e Ra te Lower-Lower Update (p=.1) Best-Best Update (p=.80) Upper-Upper Update (p=.1)
Bayes Posterior Weighting Operation P(RF), Prob.Rupture
, Failure Frequency Gi Given Failure F il , Rupture R t F Frequency Forecast: C E RC S TF F AILU RES Forecast: Rupture Rate 1.00E-02 100,000 Trial s Frequ en cy Ch art 2,647 O utliers 100,000 Trials Frequency Chart 2,056 Outliers 9.00E-03
.069 6899
.138 13787 8.00E-03 Prob abilit y D en sit y 7.00E-03
.052 .103 6.00E-03 5.00E-03
.034 Prior .069 4.00E-03 Posteri or 3.00E-03
.034
.017 2.00E-03 1.00E-03 Mean = 7.26E-6
.000 0 Mean = 1. 45E-4 0.00E+00
.000 0 0.00E+01.00E-01 2.00E-013.00E-01 4.00E-015.00E-016.00E-017.00E-01 8.00E-01 9.00E-011.00E+0 0.00E+0 1.50E-5 3.00E-5 4.50E-5 6.00E-5 0.00E+0 2.50E-4 5.00E-4 7.50E-4 1.00E-3 0 0 Fail ures per S usc eptible Wel d Year Condition al Prob abilit y of Rupture Giv en F ailur e Ruptur es per Susc epti bl e Weld Year Beliczy-Schultz Correlation and engineering 27 6/2/11 Pre-Licensing Meeting No.1 P(RF) Generic Distribution Prior judgment
Conditional Probability of Pipe Rupture vs. Break Size 1.0E+00 1.0E-01 1.0E-02 Conditional Rup pture Probability 1.0E-03 NUREG-1860 BWR Class 1 Fire Protection System Feedwater Condensate Service Water 1.0E-04 Circulating Water 1.0E-05 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
Pipe Element States S - success, no detectable damage F - detectable flaw L - detectable leak F R - rupture State Transition Rates
- flaw occurrence rate F L - leak l k ffailure il rate t
F - rupture failure rate given flaw L - rupture failure rate given leak L - repair rate via ISI exams
- repair rate via leak detection R
6/2/11 Pre-Licensing Meeting 29 No.1
Estimating Input Parameters
- Degradation related parameters estimated from service experience and Bayes models -same same as those used for the base failure rates
- Test and inspection parameters estimated using simple and easy to quantify models 30 6/2/11 Pre-Licensing Meeting No.1
Modeling g Impact p Of NDE Inspections p ((ISI))
- Capture by the repair rate for flaws PF I PF D T I TR where:
- PFI = probability that segment element with flaw will be inspected
- PFD= pprobability obab y that a flaw a is s de detected ec ed g given e inspection 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
- Capture by the repair rate for leaks PLD
( TLI TR )
where:
- PLD= probability that leak is detected given inspection
- TI = mean ea time e be between ee inspections spec o s
- TR = mean time to repair after detection 32 6/2/11 Pre-Licensing Meeting No.1
How the Model is Solved
- Model used to set up coupled differential equations of state t t
- 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 1 dr{t} 1 dr{t}
h{t}
r{t} dt 1 r{t} dt
- Hazard rate normalized to create inspection factors in pipe reliability model
- Spreadsheet is used to perform sensitivity analyses to optimize the inspection program 33 6/2/11 Pre-Licensing Meeting No.1
State Probabilities for PWR Weld subject bj to Th Thermall FFatigue i
1 0E+00 1.0E+00 1.0E-01 No Damage Probability 1.0E-02 Close to 1; STATE PRO OBABILITY Sum of all Probabilities
= 1.0 NO DETECTABLE DAMAGE 1 0E 03 1.0E-03 DETECTABLE FLAWS DETECTABLE LEAK RUPTURE 1.0E-04 1.0E-05 1 0E 06 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 2.5E-07 RUPTURE FREQ QUENCY/YEAR 2.0E-07 h{t}, Time Dependent Hazard Rate hSS; Steady State Hazard Rate 1.5E-07 40-year Average Annual Rupture Probability 1.0E-07 5.0E-08 0.0E+00 0 20 40 60 80 100 120 140 160 180 200 35 TIME IN YEARS 6/2/11 Pre-Licensing Meeting No.1
BWR Recirculation Pipe LOCA Frequency from NUREG-1860 1.0E-04 No ISI/No Leak Inspection No ISI/ Leak Inspection 1/Refueling Outage No ISI/ Leak Inspection 1/Week ISI/Leak Inspection 1/Refueling Outage BWR R Recirculation Pipin ng LOCA Frequenc cy/year 1.0E-05 ISI/Leak Inspection 1/Week 1.0E-06 1.0E-07 1.0E-08 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
- University of Maryland
- Independent review sponsored by EPRI
- 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
- 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 1.800 Change Koeberg UNIT 1 1.700 1.600 1.500 1.400 1.300 Change in CDF due to RISI 1.200 1.100 RISI No Inspection 1.000 0.900 action of CDF Acc 0.800 (Fra ceptance Criterion) 0.700 0.600 0.500 0.400 0.300 0.200 0.100 0.000 ARE ASG EAS GCT PTR RCP RCV RIS RPE RRA VVP
-0.100
-0.200 6/2/11 Pre-Licensing Meeting 38 No.1
Thermohydraulic (T/H) and Downstream Effects Analyses Overview Risk Informed GSI-191 Resolution Thursday, y, June 2,, 2011 8:00 a.m. - 5:00 p.m.
Public Meeting with STP Nuclear Operating Company Texas A&M University 6/2/11 Pre-Licensing Meeting 1 No.1
Risk Informed GSI-191 Table of Content Thermohydraulic (T/H) Simulations
- Purpose
- Outcome
- Additional Requirements
- Software
- Proposed Approach D
Downstream t Eff Effectt M Model d l (DEM)
- Purpose
- Outcome
- CFD Simulations
- RELAP5-3D Core Nodalization 6/2/11 Pre-Licensing Meeting 2 No.1
Risk Informed GSI-191 Thermohydraulic (T/H)
Simulations 6/2/11 Pre-Licensing Meeting 3 No.1
Risk Informed GSI-191 Purpose Simulations of the STP Reactor Cooling System (RCS) 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 Smallll LOCA ((< 2)
- Medium LOCA (2 ÷ 6)
- Large g LOCA ((>6))
6/2/11 Pre-Licensing Meeting 4 No.1
Risk Informed GSI-191 Outcome
- 1. Determine the values of the main thermohydraulic 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 5 No.1
Risk Informed GSI-191 Additional Requirements Analysis will be performed with:
- 1) Different Break Locations.
- 2) Different Break Sizes.
Sensitivity Analysis (SA):
- Range off C Conditions di i at which hi h Recirculation i l i is i required.
i d
- Time to Recirculation.
6/2/11 Pre-Licensing Meeting 6 No.1
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 RELAP5-3D 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, plenum core, upper plenum and downcomer [1].
[1] RELAP5-3D User Manual Volume: 1. 1 6/2/11 Pre-Licensing Meeting 7 No.1
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 DOE NNSA labs and the wider community.
It has a strongg foundation of verification and validation.
n1 n2 Dakota RELAP5 RELAP5 Output nx Output +
Input Variables Uncertainty 6/2/11 Pre-Licensing Meeting 8 No.1
Risk Informed GSI-191 Proposed Approach 1/2 STEP 1: RELAP5 Steady-State Input Certification 1A) IInputt P Preparation ti 1B) Input Certification 1C) Documentation Preparation 1D) St Steady-State d St t results lt analysis l i 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 9 No.1
Risk Informed GSI-191 Proposed Approach 2/2 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 10 No.1
Risk Informed GSI-191 STEP 1: RELAP5 Steady-State 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]
2 RELAP5-3D A 3 User Manuall Volume: l 5.
6/2/11 Pre-Licensing Meeting 11 No.1
Risk Informed GSI-191 STEP 2: LOCA Input Certification
- The input will be prepared starting from the certified steady-state input.
- SI components (A S (Accumulators, HPIS, S LPIS) S) and the system control are being i
implemented in the input.
- The input will be certified using existing simulations of LOCA (e.g., small, medium and d llarge b breakk sizes) i ) ((e.g., 3]
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 q
6/2/11 Pre-Licensing Meeting 12 No.1
Risk Informed GSI-191 STEP 3: 3D Core Nodalization Certification
- 3D Core nodalization is required for Downstream Effects Analysis 1 D Core 1-D 6/2/11 Pre-Licensing Meeting 13 No.1
Risk Informed GSI-191 STEP 3: 3D Core Nodalization Certification
- 3D Core nodalization is required for Downstream Effects Analysis 3-D 3 D Core 6/2/11 Pre-Licensing Meeting 14 No.1
Risk Informed GSI-191 STEP 4: Analysis of the Cases
- Break Locations:
Selected to account of various break locations, etc. to study the system response.
- 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 15 No.1
Risk Informed GSI-191 Downstream Effects Model (DEM) Development 6/2/11 Pre-Licensing Meeting 16 No.1
Risk Informed GSI-191 Purpose Debris may be small enough to pass through the sump screen and reach the reactor core core.
- 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 temperature distribution) will be studied.
6/2/11 Pre-Licensing Meeting 17 No.1
Risk Informed GSI-191 Outcome
- 1. Debris transport and deposition model in selected locations in the reactor vessel (downcomer, (downcomer lower plenum, fuel assemblies).
2 Simulations of the thermal hydraulics system 2.
response under conditions identified in 1 (response parameters such as fuel temperature p p distribution). )
6/2/11 Pre-Licensing Meeting 18 No.1
Risk Informed GSI-191 Software 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 19 No.1
Risk Informed GSI-191 CFD Simulations Approach
- Single Phase Flow Analysis
- Isothermal Conditions
- Discrete Phase Method (DPM) approach h tto model d l th the ddebris bi inside the reactor
- 1-way coupling Ref: Timothy D. Sande, Alionscience 6/2/11 Pre-Licensing Meeting 20 No.1
Risk Informed GSI-191 RELAP 3D C RELAP5-3D Core N Nodalization d li i Loop 1 Loop 2 Loop 4 Loop 3 6/2/11 Pre-Licensing Meeting 21 No.1
List of References
((O)) Document Available;; ((X)) Document not yyet Available
[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 1-Vol 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.
[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),
(ICTP) Trieste, Trieste Italy, Italy June 25 25-29 29, 2007.
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 Vol. 238 238, pp 2197 2197-2205 2205, 2008 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, I Aksan SN, SN OECD-LOFT Large Break LOCA Experiments: Phenomenology and Computer Code Analyses, AEEW-TRS-1003, PSI-Bericht Nr. 72, Paul Scherrer Institute, August 1990 6/2/11 Pre-Licensing Meeting 22 No.1
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, P Chataing T,T Deruaz R,R OECD/NEA/CSNI International Standard Problem No. No 2727- 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, 2 NEA/CSNI/R(92)20, 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, Test, NEA/CSNI/R(2007)4 CSNI OECD NEA, NEA October 2007.
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, 1 NEA/CSNI/R(92)20, NEA/CSNI/R(92)20 November 1992.1992
[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 23 No.1
List of References
[O] Davis D i CB, CB Assessment A t off the th RELAP5 Multi-Dimensional M lti Di i l Component C t Model M d l Using U i Data D t from f LOFT Test T t 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 Plant, ESAR-3S Pl t NUREG/CR-4384, NUREG/CR 4384 EGG-2416, EGG 2416 Idaho Id h National N ti l Engineering E i i Laboratory, L b t S t b 1985.
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 24 No.1
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, JK Bang YS, YS Kim HJ,HJ Assessment of RELAP5/MOD3.2.2gamma RELAP5/MOD3 2 2gamma with the LOFT L9 L9-3 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, 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, Evaluation, Agency 2008.
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 25 No.1