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COUPLED 3D N K AND T H TECHNIQUES AND RELEVANCE FOR THE DESIGN OF N C SYSTEMS

DIPARTIMENTO DI INGEGNERIA MECCANICA, NUCLEARE E DELLA PRODUZIONE - UNIVERSITA' DI PISA 56100 PISA - ITALY. DEPARTMENT OF MECHANICAL AND NUCLEAR ENGINEERING THE PENNSYLVANIA STATE UNIVERSITY UNIVERSITY PARK, PA 16802 - USA.

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COUPLED 3D N K AND T H TECHNIQUES AND RELEVANCE FOR THE DESIGN OF N C SYSTEMS

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  1. DIPARTIMENTO DI INGEGNERIA MECCANICA, NUCLEARE E DELLA PRODUZIONE - UNIVERSITA' DI PISA 56100 PISA - ITALY DEPARTMENT OF MECHANICAL AND NUCLEAR ENGINEERING THE PENNSYLVANIA STATE UNIVERSITY UNIVERSITY PARK, PA 16802 - USA COUPLED 3D NK AND TH TECHNIQUES AND RELEVANCE FOR THE DESIGN OF NC SYSTEMS F. D’Auria, K. Ivanov– Lecture T11 IAEA & ICTP Course on NATURAL CIRCULATION IN WATER-COOLED NUCLEAR POWER PLANTS Trieste, Italy, June 25-29 2007

  2. CONTENT • Introduction • Need for the Benchmark, • Benchmark methodology, • OECD/NEA coupled system Benchmarks • OECD/NRC PWR MSLB Benchmark, TMI-1 hypothetic transient • OECD/NRC BWR TT Benchmark, Peach Bottom-2 planned transient data • OECD/NEA/CEA V1000CT Benchmark, Kozloduy-6 planned transient data • Relevance to Natural Circulation • Conclusions

  3. INTRODUCTION Need for the Benchmark • Incorporation of a full 3D core model into system transient codes allows best-estimate simulation of interaction between core behaviour and plant dynamics • Until recently, few system codes incorporated full 3D modelling of the reactor core • For past nine years, Nuclear Energy Agency (NSC and CSNI) has developed a series of benchmarks to study the accuracy of coupled codes

  4. INTRODUCTION Need for the Benchmark The previous sets of transient benchmark problems addressed separately: • System transients (designed mainly for thermal-hydraulics codes with point kinetics models) • Core transients (designed for thermal-hydraulic core boundary conditions models coupled with a three-dimensional (3-D) neutron kinetics)

  5. INTRODUCTION Need for the Benchmark Best-Estimate Problems Plant transient benchmarks, which use a three-dimensional neutronics core model Purpose To verify the capability of system codes to analyze complex transients with coupled core/plant interactions To test fully the 3D neutronics/thermal-hydraulic coupling To evaluate discrepancies between the predictions of coupled codes in best-estimate transient simulations

  6. INTRODUCTION Benchmark Methodology • Development of thereference design from a real reactor • Definition of a benchmark problem with acomplete setof input data • Application ofthreebenchmark exercises (phases) • Evaluation ofHZP and HFP steady states • Simulation ofbest-estimate and extreme transient scenarios • Provision ofmethod for comparisonof results obtained from different codes and reference solution

  7. INTRODUCTION Benchmark Methodology Exercise OnePoint Kinetics/System Plant Simulation Exercise TwoCoupled 3D Neutronics/Thermal-Hydraulic Evaluation of Core Response Exercise ThreeBest-Estimate Coupled Core/Plant Transient Model

  8. INTRODUCTION Benchmark Methodology Any Benchmark requires a Methodology for Comparative Analysis • To evaluate discrepancies between the predictions of coupled codes in best estimate transient simulations • Different types of code results to be compared to both measured data and other code predictions • Single values, 1-D distributions, 2-D maps, and time histories • ACAP Assessment tool

  9. OECD/NEA Coupled System Benchmarks INTRODUCTION • The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently completed under the US Nuclear Regulatory Commission (NRC) sponsorship a PWR Main Steam Line Break Benchmark (MSLB) for evaluating coupled T-H system and neutron kinetics codes • A similar benchmark for codes used in analysis of a BWR plant transient has been recently defined. The NEA, OECD and US NRC have approved the BWR TT benchmark for the purpose of validating advanced system best-estimate analysis codes • VVER-1000 CT Coupled Code Benchmark Problem is a further continuation of these efforts and it defines a coupled code benchmark problem for validation of thermal-hydraulics system codes for application to Soviet-designed VVER-1000 reactors based on actual plant data

  10. OECD/NRC PWR MSLB BENCHMARK Reference Problem • Simulated Main Steam Line Break (MSLB) Break occurs in one steam line upstream of the cross-connect Control rod with maximum worth is assumed stuck out • Event is characterized bysignificant space-time effectsin the core due to the asymmetric cooling 3.Conservative assumptionsutilized to maximize RCS cool-down 4.Major concern:possible return-to-power and criticality

  11. OECD/NRC PWR MSLB BENCHMARK Evaluation of Results • Need to compare the results of over 20 different codes • Should quantify the comparison using a figure of merit • Complications • No experimental data to serve as reference calculation • Several participants submitted multiple solutions from related versions of the same code • Certain parameters are normalized so that simple averaging techniques cannot be applied

  12. OECD/NRC PWR MSLB BENCHMARK Evaluation of Results • Standard method applied for most parameters • Generate mean solution and standard deviation over all participant results for each parameter • Calculates each participant’s deviation from mean value • Divide this deviation by standard deviation to generate a figure-of-merit • Determined for each participant • Time history – at each point of interest • 2-D distribution – at each radial node • 1-D distribution – at each axial level • Normalized parameters are treated to a separate analysis to preserve normalization of mean solutions

  13. OECD/NRC PWR MSLB BENCHMARK Exercise 2 – Axial Power

  14. OECD/NRC PWR MSLB BENCHMARK Exercise Three • Combines elements of the first and second exercises and is an analysis of the transient in its entirety • Study on the impact of different NK and TH models as well as the coupling between them • Detail of spatial mesh overlays – important for local safety predictions • Modeling issues – density correlations, and spatial decay heat distribution

  15. OECD/NRC PWR MSLB BENCHMARK Exercise Three – List of participants

  16. OECD/NRC PWR MSLB BENCHMARK Exercise Three – the reference NPP & Scenario

  17. OECD/NRC PWR MSLB BENCHMARK Issues Two issues have impacted the final results of this benchmark: • Choice of Thermal-Hydraulic Model is very important for local parameters predictions during the transient (especially in the vicinity of the stuck rod) • Different Decay Heat Models have led to pronounced deviations in the transient snapshot axial power distributions after the scram

  18. OECD/NRC PWR MSLB BENCHMARK Thermal-Hydraulic Models • 18 Channel Model • These codes must lump assemblies into 18 averaged thermal-hydraulic channels ( as specified in the Specifications) • “NK Assembly” at the position of the Stuck Rod is averaged with the surrounding assemblies for the feedback modeling • The smeared in this way feedback at the stuck rod position is underestimated • 177 Channel Model - One T-H Channel (Cell) Per Assembly • Improved feedback resolution • More accurately reflects coupled behavior at stuck rod position

  19. OECD/NRC PWR MSLB BENCHMARK Reference results

  20. OECD/NRC PWR MSLB BENCHMARK Reference results

  21. OECD/NRC PWR MSLB BENCHMARK Reference results t = 10 s t = 0 s t = 90 s t = 60 s

  22. OECD/NRC PWR MSLB BENCHMARK Reference results

  23. OECD/NRC PWR MSLB BENCHMARK T-H Models – N12 Power at Return-to-Power

  24. OECD/NRC PWR MSLB BENCHMARK Decay Heat Models • To avoid uncertainties due to different models, participants were provided with: • Decay heat evolution for each scenario • Procedures to describe decay heat distributions • Average decay heat should be spatially distributed according to the initial spatial fission power distribution • This initial distribution is defined as the spatial fission power distribution at the initial HFP conditions • Deviations were still noticed: • In axial power distribution, deviations increase as time progresses after reactor scram • Some participants were re-distributing the decay heat to follow the fission power distribution at each time step

  25. OECD/NRC PWR MSLB BENCHMARK Decay Heat Models

  26. OECD/NRC BWR TT BENCHMARK • TT benchmark is established to challenge the thermal-hydraulic/neutron kinetics codes against a Peach-Bottom-2 (PB2) TT transient • Three TT transients at different power levels were performed at PB2 BWR/4 NPP prior to shutdown for refuelling at the end of Cycle 2 in April 1977 • The Turbine Trip Test 2 is chosen for the benchmark because of the impact of feedback effects and quality of the measured data

  27. OECD/NRC BWR TT BENCHMARK REF SYSTEM BIC & TSE

  28. OECD/NRC BWR TT BENCHMARK Exercise 1 • Power vs. time plant system simulation with fixed axial power profile table (obtained from the experimental data) • Purpose: To initialize and test the participants’ thermal-hydraulic system models • Core power response is fixed to reproduce the actual test results utilizing either power or reactivity vs. time data • 14 Participants have submitted their results

  29. OECD/NRC BWR TT BENCHMARK Exercise 1

  30. OECD/NRC BWR TT BENCHMARK Exercise 1

  31. OECD/NRC BWR TT BENCHMARK Exercise 1 – core average axial void fraction

  32. OECD/NRC BWR TT BENCHMARK Exercise 1 – delta dome pressure time history

  33. OECD/NRC BWR TT BENCHMARK Exercise 1 – SL pressure time history

  34. OECD/NRC BWR TT BENCHMARK Exercise 2 • Coupled 3-D kinetics/core thermal-hydraulic BC model and/or 1-D kinetics/core thermal BC model • Purpose: Qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes • Two steady states are modeled: • 1- HZP (in order to provide a clean initialization of the core neutronic models) conditions • 2- Initial condition of TT2 • 18 different results have been submitted by the participants

  35. OECD/NRC BWR TT BENCHMARK Exercise 2

  36. Average value was calculated from overall data: OECD/NRC BWR TT BENCHMARK Exercise 2 – HFP core average axial power

  37. OECD/NRC BWR TT BENCHMARK Exercise 2- HFP norm. radial power distribution

  38. OECD/NRC BWR TT BENCHMARK Exercise 2 - transient power

  39. OECD/NRC BWR TT BENCHMARK Exercise 3 • Best-estimate coupled 3-D core/thermal-hydraulic system modeling • Consists of two options: 1) 3-D core/T-H calculation for core and 1-D T-H calculation for the balance of the plant 2) 1-D kinetics core model and 1-D T-H for the reactor primary system • This exercise combines elements of the first two exercises • Also this exercise has some extreme scenarios that provide an opportunity to test better the code coupling • 15 different results have been submitted by the participants

  40. OECD/NRC BWR TT BENCHMARK Exercise 3

  41. OECD/NRC BWR TT BENCHMARK Exercise 3

  42. Axial Void Fraction 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 0 5 10 15 20 25 OECD/NRC BWR TT BENCHMARK Exercise 3 – core average axial void distribution Axial Void Fraction 0.70 0.60 0.50 0.40 PSI CEA-33 Void Fraction Void Fraction CEA-764 PUR/NRC FANP 0.30 TEPSYS FZR U.PISA GRS UPV-1 0.20 NEU NFI UPV-2 NUPEC WES. 0.10 AVERAGE AVERAGE 0.00 0 5 10 15 20 25 Axial Nodes Axial Nodes Average value was calculated from overall data excluding NEU result:

  43. OECD/NRC BWR TT BENCHMARK Exercise 3 – delta dome pressure history Measured Dome Pressure and Deviation Measured Dome Pressure and Average of the Other Data

  44. OECD/NRC BWR TT BENCHMARK Exercise 3 – TRAC-M/PARCS Results

  45. OECD/NRC BWR TT BENCHMARK Exercise 3 – TRAC-M/PARCS results

  46. OECD/NRC BWR TT BENCHMARK Exercise 3 – TRAC-M/PARCS Results

  47. OECD/NRC BWR TT BENCHMARK Exercise 3 – RELAP5/PARCS Results

  48. OECD/NRC BWR TT BENCHMARK Exercise 3 – Sample results from uncertainty estimation EVALUATION OF POWER PEAK FOLLOWING A BWR-TT EVENT AND UNCERTAINTY EVALUATION

  49. OECD/NRC BWR TT BENCHMARK Exercise 3 – Extreme Scenarios Turbine trip without bypass system relief opening Turbine trip without scram Combined Scenario – Turbine trip with bypass system relief failure without reactor scram Turbine trip with bypass system failure without scram and without safety relief valves opening SAFETY VALVES ARE MODELLED WITH 4 GROUPS

  50. OECD/NRC BWR TT BENCHMARK Exercise 3 – Extreme Scenarios TRAC-M/PARCS results TOTAL REACTIVITY POWER DOME PRESSURE VOID FRACTION

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