coupled 3d n k and t h techniques and relevance for the design of n c systems

1

COUPLED 3D COUPLED 3D NNKK AND AND TTHH TECHNIQUES TECHNIQUES AND RELEVANCE FOR THE DESIGN OF AND RELEVANCE FOR THE DESIGN OF

NNCC SYSTEMS SYSTEMS

F. D’Auria, K. Ivanov – Lecture T11

DIPARTIMENTO DI INGEGNERIA MECCANICA, NUCLEARE E DELLA PRODUZIONE - UNIVERSITA' DI PISA56100 PISA - ITALY

DEPARTMENT OF MECHANICAL AND NUCLEAR ENGINEERING THE PENNSYLVANIA STATE UNIVERSITY

UNIVERSITY PARK, PA 16802 - USA

IAEA & ICTP Course onIAEA & ICTP Course on

NATURAL CIRCULATION IN WATER-COOLED NATURAL CIRCULATION IN WATER-COOLED NUCLEAR POWER PLANTSNUCLEAR POWER PLANTS

Trieste, Italy, June 25-29 2007Trieste, Italy, June 25-29 2007

2

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

CONTENTCONTENT

3

Need for the BenchmarkNeed 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

INTRODUCTIONINTRODUCTION

4

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


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

Benchmark MethodologyBenchmark Methodology

Development of the reference design from a real reactor

Definition of a benchmark problem with a complete set of input data

Application of three benchmark exercises (phases)

Evaluation of HZP and HFP steady states

Simulation of best-estimate and extreme transient scenarios

Provision of method for comparison of results obtained from different codes and reference solution


7


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


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

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

OECD/NEA Coupled System BenchmarksOECD/NEA Coupled System Benchmarks


10

Reference ProblemReference Problem

1. 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

2. Event is characterized by significant space-time effects in the core due to the asymmetric cooling

3. Conservative assumptions utilized to maximize RCS cool-down

4. Major concern: possible return-to-power and criticality

OECD/NRC PWR MSLB BENCHMARKOECD/NRC PWR MSLB BENCHMARK

11

Evaluation of ResultsEvaluation 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

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

Evaluation of ResultsEvaluation of Results


13

ExerciseExercise 2 – Axial Power 2 – Axial Power

0.0000

0.2000

0.4000

0.6000

0.8000

1.0000

1.2000

0 5 10 15 20 25

Axial Posi tion


14

Exercise ThreeExercise 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

Exercise Three – List of participantsExercise Three – List of participantsOECD/NRC PWR MSLB BENCHMARKOECD/NRC PWR MSLB BENCHMARK

16

Exercise Three – Exercise Three – the reference NPP & Scenariothe reference NPP & Scenario


EVENT DESCRIPTION TIME (s)

Breaks open 0.0

Reactor trip 6.9

MCP trip not occurring

Turbine valve closure (start-end) 7.9-11.9

High pressure injection start 46.4

Transient end 100.0

17

IssuesIssues

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

Thermal-Hydraulic ModelsThermal-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

Reference resultsReference results


Final PK Results - Fission Power

0.00E+00

5.00E+08

1.00E+09

1.50E+09

2.00E+09

2.50E+09

3.00E+09

3.50E+09

0 2 4 6 8 10 14 18 22 26 30 34 38 42 46 50 54 58 62 66 70 74 78 82 86 90 94 98

Time (s)

Po

wer

(W

)

IPSN

Rossendorf

British Energy

Siemens/FZK

Pisa/Zagreb

VTT_1

NETCORP

GRS

Iberdrola

VTT_2

Valencia_1

Valencia_2

Purdue

PSU

20



Final PK Results - Total Reactivity

-4.50E-02

-4.00E-02

-3.50E-02

-3.00E-02

-2.50E-02

-2.00E-02

-1.50E-02

-1.00E-02

-5.00E-03

0.00E+00

5.00E-03

0 2 4 6 8 10 14 18 22 26 30 34 38 42 46 50 54 58 62 66 70 74 78 82 86 90 94 98

Time (s)

Rea

ctiv

ity

(dk/

k)

IPSN

Rossendorf

British Energy

Siemens/FZK

Pisa/Zagreb

VTT_1

NETCORP

GRS

Iberdrola

VTT_2

Valencia_1

Valencia_2

Purdue

GPUN

PSU

21



t = 0 s t = 10 s

t = 60 s t = 90 s

22

Reference resultsReference resultsOECD/NRC PWR MSLB BENCHMARKOECD/NRC PWR MSLB BENCHMARK

23

T-H Models – N12 Power at Return-to-PowerT-H Models – N12 Power at Return-to-Power

0.0000

1.0000

2.0000

3.0000

4.0000

5.0000

6.0000

7.0000

0 5 10 15 20 25

Axial Position (cm)

177 Channels

18 Channels


24

Decay Heat ModelsDecay 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

0.0000

0.4000

0.8000

1.2000

1.6000

2.0000

2.4000

1 5 9 13 17 21

Axial position

No

rmal

ized

Po

wer

ANL

BE/Tractebel 1

CEA 1

CSA/GPUN

EDF 1

EDF 2

FZR

GRS

PSU

Purdue/NRC 1

SKWU/FZK 1

SKWU/FZK 2

UP/UZ

UPM

UPV

VTT

Mean Value

Decay Heat ModelsDecay Heat Models


26

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

OECD/NRC BWR TT BENCHMARKOECD/NRC BWR TT BENCHMARK

27


REF SYSTEMBIC & TSE

28

Exercise 1Exercise 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


Name of the Participant Codes Country

CEA CATHARE V15A/Mod2.1 France

EXELON RETRAN USA

FANP S-RELAP5 V3.1.1 Germany

GRS ATHLET Germany

NETCORP DNB-3D USA

NFI TRAC-BF1 Japan

NUPEC TRAC-BF1 Japan

PSI RETRAN 3D Switzerland

PSU TRAC-BF1 USA

PSU/NRC TRAC-M USA

TEPSYS TRAC-BF1 Japan

U.PISA RELAP5/ Mod3.3 Italy

UPV TRAC-BF1 Spain

WESTINGHOUSE POLCA-T Sweden


30

Exercise 1Exercise 1 CORE AVERAGE PRESSURE DROP

0.000

0.020

0.040

0.060

0.080

0.100

0.120

0.140

0.160

PARTICIPANTS

PR

ES

SU

RE

DR

OP

(M

Pa

)

0.0125

1

2

N

xx MEASUREDi

MPaxMEASURED 113561.0

CEA EXE FANP GRS NET NFI NUP

VALUE 0.0960 0.1158 0.1130 0.1210 0.1406 0.1321 0.1041

DEVIATION -0.018 0.002 -0.001 0.007 0.027 0.019 -0.009

FOM -1.422 0.160 -0.040 0.531 1.931 1.324 -0.676

PSI PSU PSU/NRC TEP UPISA UPV WES

VALUE 0.1178 0.1130 0.1197 0.1322 0.1197 0.1159 0.1149

DEVIATION 0.004 -0.001 0.006 0.019 0.006 0.002 0.001

FOM 0.303 -0.040 0.439 1.331 0.439 0.164 0.096


31

Exercise 1 – core average axial void fractionExercise 1 – core average axial void fraction

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0 5 10 15 20 25

AXIAL NODES

VO

ID F

RA

CT

ION

CEA

EXELON

FANP

GRS

NETCORP

NFI

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.70

0 5 10 15 20 25AXIAL NODES

VO

ID F

RA

CT

ION

EXELON

NUPEC

PSI

PSU

PSU/NRC

TEPSYS

U.PISA

UPV


32

Exercise 1 – delta dome pressure time historyExercise 1 – delta dome pressure time history

0.00

0.10

0.20

0.30

0.40

0.50

0 1 2 3 4 5

time (s)

De

lta

Pre

ss

ure

Ch

an

ge

s (

MP

a)

CEA

MEASURED

FANP

GRS

NET

NFI

NUP

0.00

0.10

0.20

0.30

0.40

0.50

0 1 2 3 4 5

time (s)

MEASURED

PSI

PSU

PSU/NRC

TEP

U.PISA

UPV

WES


33

Exercise 1 – SL pressure time historyExercise 1 – SL pressure time history


34


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

Exercise 2Exercise 2PARTICIPANT CODES COUNTRY

CEA/DEN-33: CRONOS2 FLICA4 France

CEA/DEN-764: CRONOS2 FLICA4 France

FANP RAMONA5-2.1 Germany

FZR-1D: DYN3D-1D Germany

FZR-3D: DYN3D Germany

GRS QUABOX/CUBBOX-ATHLET Germany

IBERDROLA RETRAN 3D–Mod 3.1 Spain

NETCORP DNB-3D USA

NFI TRAC-BF1/COS3D Japan

NUPEC SKETCH-INS/TRAC-BF1 Japan

PSI-A RETRAN-3D MOD 003.1 Switzerland

PSI-B RETRAN-3D MOD 003.1 Switzerland

PSI-CORETRAN CORETRAN Switzerland

TEPSYS TRAC-BF1/ENTRÉE Japan

U.PISA RELAP5/ PARCS Italy

UPV TRAC-BF1/NOKIN3D Spain

VTT TRAB-3D Finland



36

Exercise 2 – HFP core average axial powerExercise 2 – HFP core average axial power

•

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0 5 10 15 20 25

Axial Nodes

No

rma

lize

d P

ow

er

CEA/DEN-33 CEA/DEN-764FANP GRSNFI NUPECPSI-A PSI-BPSI-CORETRAN VTTWESTINGHOUSE AVERAGE

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0 5 10 15 20 25

Axial Nodes

FZR-1D FZR-3D

IBERDROLA TEPSYS

U. PISA UPV

AVERAGE

Average value was calculated from overall data: N

xx i


37

Exercise 2- Exercise 2- HFP norm. radial power distribution

1 4 7 10 13 16 19 22 25 281

4

7

10

13

16

19

22

25

28

Channels

Average Normalized Radial Power, HFP

1.250-1.500

1.000-1.250

0.750-1.000

0.500-0.750

0.250-0.500

0.000-0.250

1 4 7 10 13 16 19 22 25 281

5

9

13

17

21

25

29

Channels

Deviation of Normalized Radial Power Distribution

0.40-0.450.35-0.400.30-0.350.25-0.300.20-0.250.15-0.200.10-0.150.05-0.100.00-0.05

N

xx i

1

2

N

xxi


38

Exercise 2 - tExercise 2 - transient power

0.00E+00

2.00E+09

4.00E+09

6.00E+09

8.00E+09

1.00E+10

1.20E+10

1.40E+10

1.60E+10

0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2

time (s)

Po

wer

(W)

CEA/DEN-33

CEA/DEN-764

FANP

GRS

NFI

NUPEC

PSI-A

PSI-B

PSI-CORETRAN

VTT

Average

0.00E+00

2.00E+09

4.00E+09

6.00E+09

8.00E+09

1.00E+10

1.20E+10

1.40E+10

1.60E+10

0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2

time (s)

Po

we

r (W

)

FZR-1D

FZR-3D

IBERDROLA

TEPSYS

UPV

AVERAGE


39


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 plant2) 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

Exercise 3Exercise 3PARTICIPANT CODES COUNTRY

CEA-33 CATHARE, CRONOS2, FLICA4 France

CEA-764 CATHARE, CRONOS2, FLICA4 France

FANP S-RELAP5 / RAMONA5-2.1 Germany

FZR DYN3D-ATHLET Germany

GRS ATHLET-QUABOX/CUBBOX Germany

NEU THYDE-NEU Japan

NFI TRAC-BF1/COS3D Japan

NUPEC SKETCH-INS/TRAC-BF1 Japan

PSI RETRAN-3D MOD 003.1 Switzerland

PURDUE/NRC TRAC-M/PARCS USA

TEPSYS TRAC/BF1-ENTRÉE Japan

U.PISA RELAP5/PARCS Italy

UPV-1 (MODKIN) TRAC-BF1;MODKIN Spain

UPV-2 (NOKIN-3D) TRAC-BF1;NOKIN-3D Spain



41

•

Exercise 3Exercise 3Keff

0.985

0.990

0.995

1.000

1.005

1.010

1.015

Participants

Keff

004789.1N

xx i


42

•

Exercise 3 – core average axial void distributionExercise 3 – core average axial void distributionAxial Void Fraction

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.70

0 5 10 15 20 25

Axial Nodes

Vo

id F

ract

ion

CEA-33

CEA-764

FANP

FZR

GRS

NEU

NFI

NUPEC

AVERAGE

Average value was calculated from overall data excluding NEU result:

Axial Void Fraction

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.70

0 5 10 15 20 25

Axial Nodes

Vo

id F

ract

ion

PSI

PUR/NRC

TEPSYS

U.PISA

UPV-1

UPV-2

WES.

AVERAGE

N

xx i


43

Exercise 3 – delta dome pressure history Exercise 3 – delta dome pressure history

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0

time (s)

Del

ta P

ress

ure

Ch

ang

es (

MP

a)

(+) deviation

MEASURED

(-) deviation

0.00

0.10

0.20

0.30

0.40

0.50

0.60

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0

time (s)

Del

ta P

ress

ure

Ch

ang

es (

MP

a)

MEASURED

AVERAGE

Measured Dome Pressure and Deviation

Measured Dome Pressure and Average of the Other Data

1

2

N

xxiN

xx i


44

Exercise 3 – TRAC-M/PARCS ResultsExercise 3 – TRAC-M/PARCS Results

Steady State Power DistributionPBTT Exercise 3

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0.0 0.2 0.4 0.6 0.8 1.0

Relative Height

Rel

ativ

e Po

wer

Measurement

TRACM/PARCS


45

Exercise 3 – TRAC-M/PARCS resultsExercise 3 – TRAC-M/PARCS results

Steady State Void ProfilePBTT Exercise 3

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Relative Height

Voi

d Fr

actio

n

TRACM/PARCS

RETRAN1D


46

Exercise 3 – TRAC-M/PARCS ResultsExercise 3 – TRAC-M/PARCS Results

Total Power

PBTT Exercise 3

0

50

100

150

200

250

300

0 0.2 0.4 0.6 0.8 1 1.2 1.4

Time (s)

Po

we

r (%

)

TRACM/PARCS

Measurement


47

Exercise 3 – RELAP5/PARCS ResultsExercise 3 – RELAP5/PARCS Results


48

Exercise 3 –Exercise 3 – Sample results from uncertainty estimation Sample results from uncertainty estimation


EVALUATION OFEVALUATION OFPOWER PEAKPOWER PEAKFOLLOWING FOLLOWING A BWR-TT EVENT A BWR-TT EVENT AND UNCERTAINTY AND UNCERTAINTY EVALUATION EVALUATION

49

Exercise 3 – Extreme ScenariosExercise 3 – Extreme Scenarios

1) Turbine trip without

bypass system relief opening

2) Turbine trip without scram

3) Combined Scenario – Turbine trip with bypass system relief failure without reactor scram

4) Turbine trip with bypass system failure without scram and without safety relief valves opening

Turb ine

C O N D E N SE R

CO NTRO LRO DS TU R B IN E

C O N TRO LVA LV E

TU R B IN EB YPA SSVA LV E

TU R B IN E S TO PVA LV E

RV S V H C P I

M S IV

P

H P C I

FE EDW ATE R

C V

SAFETY VALVES ARE MODELLEDSAFETY VALVES ARE MODELLED

WITH 4 GROUPSWITH 4 GROUPS


50

PBTT2 Benchmark Exercise 3 Extreme Scenarios 1, 2, 3 and 4 Comparison

TRAC-M/PARCS

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

4.5

10.0 11.0 12.0 13.0 14.0 15.0 16.0 17.0 18.0 19.0 20.0

time (sec)

Rat

ed P

ow

er

Extr 4

Extr 3

Extr 2

Extr 1

-> After 10 sec null transient


TRAC-M/PARCS

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

1.0

10.0 11.0 12.0 13.0 14.0 15.0 16.0 17.0 18.0 19.0 20.0

time (sec)

To

tal R

eac

tiv

ity

Extr 4

Extr 3

Extr 2

Extr 1



TRAC-M/PARCS

6.50E+06

7.50E+06

8.50E+06

9.50E+06

1.05E+07

1.15E+07

1.25E+07

1.35E+07

10.0 11.0 12.0 13.0 14.0 15.0 16.0 17.0 18.0 19.0 20.0

time (sec)

Do

me

Pre

ss

ure

(P

a)

Extr 1

Extr 2

Extr 3

Extr 4



TRAC-M/PARCS

0.00

0.05

0.10

0.15

0.20

0.25

0.30

0.35

10.0 11.0 12.0 13.0 14.0 15.0 16.0 17.0 18.0 19.0 20.0

time (sec)

Vo

id F

rac

tio

n

Extr 1

Extr 2

Extr 3

Extr 4


Exercise 3 – Extreme Scenarios TRAC-M/PARCS resultsExercise 3 – Extreme Scenarios TRAC-M/PARCS results

POWER TOTAL REACTIVITY

DOME PRESSURE

VOID FRACTION


51

Extreme scenario # 3 - Extreme scenario # 3 - RV and SRV DISCHARGE RV and SRV DISCHARGE

109876543210

500.0

450.0

400.0

350.0

300.0

250.0

200.0

150.0

100.0

50.0

0.0


52

Exercise 3 – Extreme scenario 4Exercise 3 – Extreme scenario 4

Extreme scenario #4 - WITHOUT BYPASS, WITHOUT

SCRAM and WITHOUT ACTIVATION of SRVs

This scenario is an example to compare physical models without external perturbations (no need of modeling of SRVs and their locations), possibility to determine the eigen-frequence of the system


53

The reference power plant design for this analysis is Unit 6 at the Kozloduy NPP site in Bulgaria. This plant is a VVER-1000 Model V320 pressurized water reactor that produces 3000 MW thermal power and generates 1000 MW electric power

During the plant-commissioning phase at Kozloduy NPP – Unit #6 a number of experiments were performed

One of them is the investigation of the behavior of the nuclear power reactor parameters in case of switching on one main coolant pump (MCP) when the other three main coolant pumps are in operation

OECD/DOE/CEA VVER-1000 CT BenchmarkOECD/DOE/CEA VVER-1000 CT Benchmark

54

Transient ScenarioTransient Scenario

The transient test scenario is as follows:

At reactor power 29.45% Nnom MCP#3 is switched on

After switching on MCP#3 the reactor power increases to 29.8%Nnom

Pressurizer water level decreases from 744 cm to 728cm

Water level in the Steam Generator #3 decreases with 9 cm

The flow rate in loop #3 reverses back to normal at the 13th sec. of the switching on MCP#3. The timing is consistent with reactivity increase, as observed through the reactor power set points


55

In addition, an extreme version is defined as follows:

The control rod of group 10 located in the sector of the core cooled by MCP 3 is ejected after switching on of the MCP 3

The scram is activated upon reaching the high neutron flux set point, which is 110% of the initial level

Advantages:

This extreme scenario will develop very peaked spatial power distribution and nonlinear asymmetric feedback effects

It is designed to test and compare better the predictions of coupled 3-D kinetics / thermal-hydraulic codes

Extreme ScenarioExtreme Scenario


56

Transient ScenarioTransient Scenario

1 2 3 4 5 6

7 8 9 IV

10 11 IX

12

16 17 III

18 19 VIII

20 II

21 I

13 III

14 15

22 VII

23 24 IV

25

61

26 27 28 VII

29 30 31 VI

32 33 34 VIII

35 36

37 38 IX

39 I

40 41 X

42 43 44 X

45 46 II

47 IX

48

49 50 51 II

52 V

53 54 55 VI

56 57 58 V

59 I

60

62 63 IV

64 VIII

65 66 67 VI

68 69 70 VI

71 72 73 VII

74 III

75

101 IV

102

76 77 78 79 X

80 81 82 V

83 84 85 X

86 87

89 90 III

91 VII

92 93 94 VI

95 96 97 VI

98 99 100 VIII

115 103 104 105 I

106 VI

107 108 109 VI

110 111 112 VI

113 II

114

116 117 IX

118 II

119 120 X

121 122 123 X

124 125 I

126 IX

127

128 129 130 VIII

131 132 133 V

134 135 136 VII

137 138

139 140 IV

141 142 VII

143 I

144 II

145 VIII

146 147 III

148

149 150 151 III

152 153 IX

154 155 IV

156 157

158 159 160 161 162 163

88

I

II

III

IV

20° 30’

34° 30’

34°30’

20°30’

MCP 1

MCP 2

MCP 3

MCP 4

A B

FA with Control Rod A – FA number B – Control Rod Group number

I, II, III, IV – Reactor Vessel and Core Axes


57

Core and Neutronic DataCore and Neutronic Data

1 1.219

2 1.283

3 1.311

4 1.553

5 1.360

6 1.409

7 1.391

29 0.0

8 1.603

9 1.271

10 1.286

11 1.676

12 1.578

13 0.909

29 0.0

14 1.271

15 1.213

16 1.271

17 1.246

18 1.065

29 0.0

19 1.286

20 1.271

21 1.489

22 1.053

29 0.0

23 1.676

24 1.246

25 1.053

29 0.0

26 1.578

27 1.065

29 0.0

28 0.909

29 0.0

29 0.0

1

2

2-D assembly type map


Fuel assembly with enrichment 2.0%



Profiled fuel assembly with enrichment 3.3%

Reflector Assembly

1 – Type of fuel assembly2 – Burnup MWd/kgU

58

In summary the transient was chosen because of the following reasons:

It is a real transient of an operating VVER-1000 power plant

Sufficient experimental data exists – there is measured data for the initial conditions with three working MCP as well as during the transient following the switching on of MCP #3

There is a rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatial distributed positive reactivity due to the modeled feedback mechanisms

During the transient there is a non - symmetric power distribution


59

V1000CT-1 DescriptionV1000CT-1 Description

Phase 1 of the benchmark is based on a comparison with the NPP experiment of a MCP start up when the other three pumps are in operation

This thermal-hydraulic driven transient is characterized by spatially dependant non-symmetric processes. The NK asymmetry is enforced in the added extreme scenario of Exercise 3 – Rod ejection

Three exercises are defined in order to verify the capability of system codes to analyze complex transients with coupled core‑plant interactions and to test fully the 3-D NK/TH coupling

60

V1000CT-1 Exercise 1 – Point kinetics plant simulation The purpose of this exercise is to test the primary and secondary system model responses and initialize the system model.

V1000CT-1 Exercise 2 – Coupled 3-D neutronics/core thermal-hydraulics response evaluationThe purpose of this exercise is to model the core and the vessel only.

V1000CT-1 Exercise 3 – Best-estimate coupled code plant transient modelingIn this exercise the participants must analyze the transient in its entirety, and computation results will be compared to measured plant data.


61

Since previous benchmarks indicate that further improvement of the mixing computation models in the integrated codes is necessary, a coolant mixing and MSLB benchmark for VVER-1000 was defined in phase 2 of the benchmark (V1000CT-2)

V1000CT-2 Exercise 1: Computation of coolant mixing experiments

This exercise is based on a comparison with a mixing experiment conducted at Kozloduy-6 as part of the plant-commissioning phase

The experiment includes isolation of a steam generator at 9.3% of the nominal power causing single loop heat-up, with all MCP in operation


62

V1000CT-2 Exercises 2 and 3: Main Steam-Line Break (MSLB) modeling

The transient to be analyzed is initiated by a main steam line break in a VVER-1000 between the steam generator (SG) and the steam isolation valve (SIV), outside the containment

This event is characterized by a large asymmetric cooling of the core, stuck control rods and a large primary coolant flow variation

Two scenarios will be defined: the first scenario is taken from the current licensing practice and the second is derived from the original one using aggravating assumptions to enhance the code-to-code comparison


63

The main objective of the study is to clarify the local 3-D feedback effects depending on the vessel mixing

Special emphasis is put on testing 3-D vessel thermal-hydraulic (T-H) models and the coupling of 3-D neutronics/vessel thermal hydraulics

The MSLB is thus divided in two exercises (to be done for the two scenarios):

Exercise 2 consists of coupled 3-D neutronics/vessel thermal-hydraulic simulations using specified vessel T-H boundary conditions and

Exercise 3 consists of best estimate coupled plant simulations (plant, 3-D vessel and core)


64

V1000CT-1 : Penn StateExercise 1 : point kinetics plant simulation

Exercise 2 : coupled 3D neutronics/core TH response

Exercise 3 : best-estimate coupled core/plant simulation

V1000CT-2 : CEA-INRNEExercise 1 : calculation of NPP coolant mixing experiments (temperature and flow variation)

Exercise 2 : MSLB, coupled 3D neutronics/3D vessel TH simulation

Exercise 3 : MSLB best-estimate coupled core/plant modeling


65

Organizer: OECD/NEA (NSC & CSNI)

Sponsors: US-DOE and CEA

Coordinators: Penn State, INRNE and CEA

Collaborations:–AER Working Group D–KNPP (Kozloduy-6 data) and INRNE–ANL


66

Participants’ information

Name Country Code

CEA France CATHAREFZK Germany RELAP5/MOD3.3GRS Germany ATHLETINRNE Bulgaria RELAP5/MOD3.2KU Ukraine RELAP5-3DNRI Czech Republic ATHLET 1.2 Cycle DORNL USA RELAP5-3DPSU USA TRAC-PF1/MOD2UPISA Italy RELAP5/MOD3.3


67

TRANSIENT RESULTS – Exercise 1TRANSIENT RESULTS – Exercise 1

Total power change

0.00E+00

1.00E+07

2.00E+07

3.00E+07

4.00E+07

5.00E+07

6.00E+07

0 200 400 600 800

Time, s

De

lta

po

we

r c

ha

ng

e, W

68

TRANSIENT RESULTS – Exercise 1TRANSIENT RESULTS – Exercise 1

Hot leg 3 coolant temperature change

-2.00

0.00

2.00

4.00

6.00

8.00

10.00

12.00

14.00

16.00

18.00

0 20 40 60 80 100 120

Time, s

De

lta

te

mp

era

ture

ch

an

ge

, K

Participant FOMFZK 0.849INRNE 0.870KU 0.845NRI 0.840ORNL 0.843PSU 0.840UPISA 0.828

69

SS (Basic scenario) – Exercise 3SS (Basic scenario) – Exercise 3

HP SS Normalized axial power distribution

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

0 2 4 6 8 10 12

Axial node

No

rmal

ized

po

wer

FZR

KU

PSU

UPISA

70

TR (Basic scenario) – Exercise 3TR (Basic scenario) – Exercise 3

Total power change

0.00E+00

5.00E+06

1.00E+07

1.50E+07

2.00E+07

2.50E+07

3.00E+07

3.50E+07

4.00E+07

4.50E+07

0 100 200 300 400 500 600 700 800

Time, s

Del

ta p

ow

er c

han

ge,

W

71

Cold leg 3 coolant temperature change

-4.50

-4.00

-3.50

-3.00

-2.50

-2.00

-1.50

-1.00

-0.50

0.00

0.50

1.00

0 20 40 60 80 100 120

Time, s

Del

ta t

emp

erat

ure

ch

ang

e, K

TR (Basic scenario) – Exercise 3TR (Basic scenario) – Exercise 3

72

TR (Extreme scenario) – Exercise 3TR (Extreme scenario) – Exercise 3

Extreme scenarioTotal power change

0.00E+00

5.00E+07

1.00E+08

1.50E+08

2.00E+08

2.50E+08

0 100 200 300 400 500 600 700 800

Time, s

Del

ta p

ow

er c

han

ge,

W

73

Core average density change

0.00

0.50

1.00

1.50

2.00

2.50

3.00

3.50

4.00

4.50

5.00

0 100 200 300 400 500 600 700 800

Time, s

Del

ta d

ensi

ty c

han

ge,

kg

/m3

Initial value

FZR 7.52E+02

KU 7.48E+02

PSU 7.49E+02


74

Core average fuel temperature change

-5.00

0.00

5.00

10.00

15.00

20.00

25.00

30.00

0 100 200 300 400 500 600 700 800

Time, s

Del

ta t

emp

erat

ure

ch

ang

e, K

Initial value

FZR 6.64E+02

KU 6.61E+02

PSU 6.77E+02


75

RELEVANCE TO NATURAL CIRCULATION

Working conditions for existing NPP imply:

o NK-TH feedback,o Operation of MCP (apart a few exceptions).

NC and NK-TH feedback are ‘irrelevant’ for NPP nominal operation

Areas where NC and coupled 3D NK-TH techniques become relevantAreas where NC and coupled 3D NK-TH techniques become relevantincludeinclude

Design of innovative reactors,Design of innovative reactors, BWR stability (special lecture in this course),BWR stability (special lecture in this course), ATWS and selected RIA (e.g. CR expulsion at HZP), ATWS and selected RIA (e.g. CR expulsion at HZP), Boron dilution transients. Boron dilution transients.

No specific effort has been made so far to connect NC and the use of No specific effort has been made so far to connect NC and the use of coupled 3D NK-TH techniques (apart from the area of BWRS). coupled 3D NK-TH techniques (apart from the area of BWRS).

76

CONCLUSIONSCONCLUSIONS - BENCHMARK ACTIVITIES - BENCHMARK ACTIVITIES

PWR MSLB Benchmark is completed: 3 OECD/NEA reports were published, and special issue of Nuclear Technology was published

BWR TT Benchmark is completed: 3 OECD/NEA reports are being prepared, and special issue of Nuclear Engineering is underway

V1000CT Benchmark is being currently conducted

77

CONCLUSIONS CONCLUSIONS -- CURRENT STATUS CURRENT STATUS IN THE APPLICATION OF 3D NK-TH TECHNIQUESIN THE APPLICATION OF 3D NK-TH TECHNIQUES

> CAPABILITIES OF 3D TH - NK TOOLS CHARACTERIZEDCAPABILITIES OF 3D TH - NK TOOLS CHARACTERIZED

> 3D ANALYSES OF SELECTED SCENARIOS RECOMMENDED 3D ANALYSES OF SELECTED SCENARIOS RECOMMENDED

> THRESHOLDS OF ACCEPTABILITY PROPOSED THRESHOLDS OF ACCEPTABILITY PROPOSED

> RELEVANCE OF THE ROLE OF FUEL IDENTIFIEDRELEVANCE OF THE ROLE OF FUEL IDENTIFIED

> UNCERTAINTY SOURCES (MAIN) IDENTIFIED:UNCERTAINTY SOURCES (MAIN) IDENTIFIED:- Sub-cooled HTC

- Pressure wave propagation inside RPV

- Direct energy release to the coolant including (prompt) radiolysis

- Specifications for components such as valves and control rods.

- Time functions for fuel related parameters

- NK parameters (CR worth, β and Doppler known with uncertainties given by 10%, 5% and 20%).

> LICENSING STATUSLICENSING STATUS- 3D TH-NK RIA & ATWS analyses shall become mandatory

> A DATABASE OF INPUT AND OUTPUT CREATED A DATABASE OF INPUT AND OUTPUT CREATED

78

CONCLUSIONS CONCLUSIONS -- CURRENT STATUS CURRENT STATUS IN THE APPLICATION OF 3D NK-TH TECHNIQUESIN THE APPLICATION OF 3D NK-TH TECHNIQUES

> SYSTEMATIC QUALIFICATION OF INDIVIDUAL STEPS OF SYSTEMATIC QUALIFICATION OF INDIVIDUAL STEPS OF THE PROCESSTHE PROCESS

> FULL CONSIDERATION TO THE IDENTIFIED SOURCES OF FULL CONSIDERATION TO THE IDENTIFIED SOURCES OF UNCERTAINTIESUNCERTAINTIES

> RESULTS FROM 0-D TH-NK ARE NOT CONSERVATIVERESULTS FROM 0-D TH-NK ARE NOT CONSERVATIVE

> NUCLEAR FUEL RELATED MODELS SHALL BE INTEGRATED NUCLEAR FUEL RELATED MODELS SHALL BE INTEGRATED INTO TH-NK COUPLING (THIS ALSO INCLUDES THE INTO TH-NK COUPLING (THIS ALSO INCLUDES THE CHEMISTRY, E.G. THE PROCESS OF CRUD FORMATION AND CHEMISTRY, E.G. THE PROCESS OF CRUD FORMATION AND RELEASE)RELEASE)

> THE INDUSTRY AND THE REGULATORY BODIES SHOULD THE INDUSTRY AND THE REGULATORY BODIES SHOULD BECOME FULLY AWARE OF THE CAPABILITIES (AND OF BECOME FULLY AWARE OF THE CAPABILITIES (AND OF THE LIMITATIONS) OF THE CONCERNED TECHNIQUESTHE LIMITATIONS) OF THE CONCERNED TECHNIQUES

NEW FRONTIER

coupled 3d n k and t h techniques and relevance for the design of n c systems

Documents

system transient codes

predictions of coupled

benchmark problem

similar benchmark

capability of system

different codes

d core model

system transients