w. bernnat, m. buck, n. bensaid, ike, university of stuttgart · w. bernnat, m. buck, n. bensaid,...

The PBMR-400 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux - 26-27 January 2006

Universität Stuttgart

Institut für Kernenergetik und Energiesysteme

Application and development oftools for HTR neutronics and

thermal hydraulics analysis at IKE

ApplicationApplication and and development development of oftools fortools for HTR neutronics and HTR neutronics and

thermalthermal hydraulics analysis hydraulics analysis at IKE at IKEW. Bernnat, M. Buck, N. BenSaid,

K. Hossain, M. MesinaIKE,

University of Stuttgart

The PBMR-400 Core Design - 2nd Workshop - OECD/NEA, Issy les Moulineaux - 26-27 January 2006 2



General Objectives: Development of an Integrated Tool forthe Analysis of Stationary and Dynamic Behavior of HTRs

●Neutronics reactordesign (stationaryand transient).

●In-CoreThermal-hydraulics andfuel thermalbehavior.

●Modeling ofcomponents anddynamics ofPower ConversionUnit (PCU).

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HTR applications

• Neutronics stationary• Neutronics transient• Thermal hydraulics• PCU simulation




Codes available• ZIRKUS , KIND• THERMIX/KONVEK• ANISN/DORT/TORT (1D-3D SN-transport)• Monte Carlo : MCNP(X)/KENO• NJOY (Cross section preparation)• RSYST (IKE), general reactor physics applications• MICROX-2• RESMOD (IKE), multicell spectralcode• FLOWNEX (M-Tech), system code




ZIRKUS• Modular system with funcional modules and data base system• Restricted to pebble bed HTR types• Flexibility high, but rather complex• Several adaptations to PBMR like systems (annular core) were made together with simplifications of input structure




Main modular ZIRKUS sequence for stationaryneutronics - thermal hydraulics calculations

KUG EL Defin ition of spherical fuel e lem entW Q RFL Cross-sections for reflectors (tem perature,

density, im purity)N IVERM Num ber densities in itia lisation/shufflingNEW A Dancoff factorsM ICRO X Cross section resonance treatm ent and

spectral ca lculation (m icroscopic XS forspectral zones)

M AG RU M acroscopic cross sections for burnup zonesHBLO CK 2D diffusion calculation, variable group

num bers, Xe distributionVO RNEK Power density ca lculation




Main modular ZIRKUS sequence for stationaryneutronics - thermal hydraulics calculations

Z B U C K B u c k l in g s f o r s p e c t r a l z o n e s d e r i v e df r o m d i f f u s io n c a lc u la t io n

S B U R N D e p le t io n c a lc u la t io n f o r a l l b u r n u pz o n e s

N Z W D e c a y h e a t c a lc u la t io nN Z W A U S In t e r f a c e d e c a y h e a t d a t aL D T H E R M I In t e r f a c e Z IR K U S - T H E R M IXT H E R M IX T h e r m a l h y d r a u l ic c a lc u la t io nT Z IR K In t e r f a c e T H E R M IX - Z IR K U SD IF K D i f f u s io n c o n s t a n t s f o r c a v i t y r e g io n




ZIRKUS-features• first core• transition core• equilibrium core• flow pattern of pebbles• reload strategy• fuel/moderator elements• 2 D representation for burn up• 3 D stationary calculations (burnup distribution 2D) with

Monte-Carlo codes MCNP and/or KENO• temperature coefficients• water ingress• xenon reactivity• flexibility, connection to other codes possible



Institut für Kernenergetik und EnergiesystemeTypical ZIRKUS/THERMIX applications

Cross-section

DatabaseMICROX/

MCNP

FIRST-CORETRANSITION COREEQUILIBRIUM CORE

Calculation ofReactivityCoefficients.Database forTransients MonteCarlo, SN-Transport

ZIRKUSDatabase

Thermal- LOFC hydraulic DLOCA analysis




ZIRKUS options

SINGLE MODULE INPUTS

SEQUENCE OF MODULES

EXECUTION OF SINGLE MODULES AND SEQUENCES

ARCHIVING OF CALCULATIONAL RESULTS

RESTART-OPTION

INTERFACE TO THERMAL HYDRAULICS

INTERFACE TO TRANSPORT PROGRAMS

INTERFACE TO TRANSIENT CODES KIND; RZ KIND



Institut für Kernenergetik und EnergiesystemeZIRKUS modular chain

HBLOCK

KUGEL, DIFK

WQRFL

NIVERM

VORNEK

MAGRU

MICROX

NEWA

ZBUCK

SBURN

NZW

NZWAUS

LDTHERMI

THERMIX

TZIRK



Institut für Kernenergetik und EnergiesystemeSimulation platform




Cross section data base and spectral code

• Data evaluations: JEFF 3.1• MICROX-2 spectral code• RESMOD spectral code• MCNP(X) Monte Carlo• Thermal neutron scattering laws for graphite




MICROX-2 for detailed calculation of spectra in fast,resonance and thermal range

FDTAPE92 groupsfast range

GARTAPEPointwiseResolved

resonan-cerange

GGTAPE101

groupsthermalrange

Geometryof cell,

temperatureDancofffactor,

buckling

ZonewiseNuclide

composi-tion

MICROX-2

Condensedmicroscopic

crosssections

Z

I

R

K

U

S



Institut für Kernenergetik und EnergiesystemePresent MICROX-2 Library based on JEFF-3.1

158 Nuclides

1-H-1 34-Se-79 47-Ag-109 58-Ce-141 63-Eu-151 90-Th-232 95-Am-241 5-B-10 36-Kr-83 47-Ag-110m 58-Ce-142 63-Eu-153 91-Pa-233 95-Am-242 5-B-11 37-Rb-87 48-Cd-110 58-Ce-143 63-Eu-154 93-Np-235 95-Am-242m 6-C-0 38-Sr-90 48-Cd-111 58-Ce-144 63-Eu-155 93-Np-236 95-Am-243 7-N-14 40-Zr-93 48-Cd-113 59-Pr-141 63-Eu-156 93-Np-237 95-Am-244 8-O-16 40-Zr-95 50-Sn-126 59-Pr-143 63-Eu-157 93-Np-238 95-Am-244m14-Si-28 41-Nb-94 52-Te-129m 60-Nd-142 64-Gd-152 93-Np-239 96-Cm-24014-Si-29 41-Nb-95 52-Te-132 60-Nd-143 64-Gd-154 92-U-233 96-Cm-24114-Si-30 42-Mo-100 53-I-129 60-Nd-144 64-Gd-155 92-U-234 96-Cm-24224-Cr-50 42-Mo-95 53-I-131 60-Nd-145 64-Gd-156 92-U-235 96-Cm-24324-Cr-52 42-Mo-99 53-I-135 60-Nd-146 64-Gd-157 92-U-236 96-Cm-24424-Cr-53 43-Tc-99 54-Xe-131 60-Nd-147 64-Gd-158 92-U-237 96-Cm-24524-Cr-54 44-Ru-101 54-Xe-132 60-Nd-148 64-Gd-160 92-U-238 96-Cm-24625-Mn-55 44-Ru-102 54-Xe-133 60-Nd-150 65-Tb-159 94-Pu-238 96-Cm-24726-Fe-54 44-Ru-103 54-Xe-134 61-Pm-147 65-Tb-160 94-Pu-239 96-Cm-24826-Fe-56 44-Ru-104 54-Xe-135 61-Pm-148 66-Dy-160 94-Pu-240 96-Cm-24926-Fe-57 44-Ru-106 55-Cs-133 61-Pm-148m 66-Dy-161 94-Pu-241 96-Cm-25026-Fe-58 45-Rh-103 55-Cs-134 61-Pm-149 66-Dy-162 94-Pu-24228-Ni-58 45-Rh-105 55-Cs-135 61-Pm-151 66-Dy-163 94-Pu-24328-Ni-60 46-Pd-104 55-Cs-137 62-Sm-147 66-Dy-164 94-Pu-24428-Ni-61 46-Pd-105 57-La-139 62-Sm-148 67-Ho-16528-Ni-62 46-Pd-106 62-Sm-149 72-Hf-17628-Ni-64 46-Pd-107 62-Sm-150 72-Hf-177

46-Pd-108 62-Sm-151 72-Hf-17862-Sm-152 72-Hf-17962-Sm-153 72-Hf-18062-Sm-154

Temperatures: 300; 600; 800; 900; 1100; 1400; 1800; 2400 K



Institut für Kernenergetik und EnergiesystemeThermal neutron scattering on graphite

Frequency distributions of graphite

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5

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35

40

0.00 0.05 0.10 0.15 0.20 0.25Omega in eV

Rho

(Om

ega)

in 1

/eV

GA

ORNL

NCSU




Comparison of measured and calculated spectra

0.0

1.0

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5.0

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Neutron Energy (eV)

EPhi

(E) (

arbi

trary

uni

ts)

Experiment GAC-5A 274 K

GAC-5a : GA modelGAC-5a : ORNL model

GAC-5a : NCSU modelExperiment GAC-5F 600 K

GAC-5f : GA modelGAC-5F : ORNL model




Transient calculations

⇒ ZKIND and ⇒ RZKIND codes

Both are neutron kinetic/dynamic codes for the pebble bed core primarily developed for the HTR-MODUL reactor.

The Tasks are: calculation of

• time dependent power distributions and reactivity effects• temperature distributions




Thermal-hydraulic Models

Different models to simulate the heat productionand temperature distribution in fuel

• Macro Model (2D RZKIND)• Macro and Micro Model (1D ZKIND)• enhanced Micro Model (1D PKIND)




Macro and Micro modelof Heat conduction

Macro Model:

In RZKIND at the time a homogeneous model is implemented. TheFuel spheres are subdivided into several shells but the fueltemperature is assumed to be identical to the graphite temperature inthe corresponding shell. The fuel temperature calculated forreactivity feedback is averaged over all shells containing fuel (exceptouter shell). The moderator temperature is the average of graphitetemperatures of all shells.




Thermal-hydraulic Modelsused in KIND codes

Macrosystem Microsystem

ShellMatrix

Kernel

Coating

Share of MatrixGraphite

τ =T -

τ : Temperature increase in the particle : average Temperature over the fine structure

rM

rS

rK

rC

rM M




Data Flow to RZKIND

Data Library User’s InputFile

RZKIND

ZIRKUS

OUTPUTFILE

ZIRKUSGeometryCross Section data

KONTRCondensation of cross sections

ZIRKUS-Geometry RZKIND-Geometry

ZKIND

VPFPolynomial fit of cross sections

RZKINDData-Library




Features of the KIND CodesSimulate• Inlet temperature disturbances• Mass flow disturbances• Changes in power• Reactivity disturbances

•Control rod movement or SAS insertion•Xenon effects•external reactivity effects




Example: Control rod withdrawal(2D RZ-KIND calculation for HTR Modul)

(a) first scram works at 120% neutron flux

(b) second scram works if average He coolant outlet Temperature exceeds limit

(c) neither 1. Scram nor 2. scram works control rods withdraw to the maximum upper end position (-100cm)

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(a)

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control rod




Withdrawal of all rods with 100 cm/s

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RZKIND (bue)calculation

Particle Modell ofZKIND

Influence of detailed fuel temperature model(control rod velocity: 100 cm/s)

(blue) 2D RZ-KIND calculation

(red) 1D ZKIND particle model calculation

Extreme differences betweenthe two models




Fast transient calculation

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FuelCoat.1Coat.2Matrix-max

Matrix-averageFE_Shell

Fuel Element Shell

average Matrix Temperature

UO2 TemperatureZKIND (Particle model)

- 100 cm/s




General Objectives :

• Reliable and flexible integrated tool for analysis of static anddynamic behavior of HTRs:

• Emphasis on detailed description of in-vessel behavior,especially coupling of thermal-hydraulics and neutronics

• Existing tools as starting point for further development

• Present code THERMIX

Thermal hydraulics




HTR with annular core [PBMR]THERMIX discretisationce

ntra

l gra

hite

col

umn

annu

lar c

ore

Initial steady state temperature distribution

Design of PBMR with annular core andcompact central column• Thermal power: 400 MW

• System pressure: 85 bar

• Inlet temperature: 500 °C

• Outlet temperature: 900 °C

Examples of analyses and applications




Results for Annular Core ReactorLOFC with depressurization LOFC without depressurization:

Gas temperature and velocity(annular core only)

Solid temperature development Solid temperature development

Examples of analyses and applications




Further developments• 3D calculations in ZIRKUS• Improved treatment of streaming in cavities for

diffusion calculations• New 2D/3D thermal hydraulics module• Extensions of the space time kinetics modules

– more energy groups– more flexible mesh grid– coupling to new thermal hydraulics code– detailed heat transfer model for coated particles

embedded into graphite matrix for 2D version




Benchmark PBMR 400Neutronics/Thermal hydraulics

• ZIRKUS and DORT model (exercise 1)• THERMIX and KONVEK model (exercise 2)



Institut für Kernenergetik und EnergiesystemeMesh and media assignment : Steady State Exercise 1

73,6 80,55 92,05 100 117 134 151 168 185 192,5 204,5 211,4 225 243,6 260,6 275 287,5 292,5Mesh subdivision

0 10 41 73,6 80,6 92,1 100 117 134 151 168 185 193 204 211 225 244 261 275 288 2931 2 3 1 1 1 2 2 2 2 2 1 1 1 2 2 2 2 2 1

-200 10 31 32,6 6,95 11,5 7,95 17 17 17 17 17 7,95 11,5 6,95 13,6 18,6 17 14,4 12,5 5-150 5 50 133 133 133 133 155 116 113 113 113 113 113 135 164 144 144 152 152 152 189 190-100 5 50 133 133 133 133 155 116 113 113 113 113 113 135 164 144 144 152 152 152 189 190-50 5 50 133 133 133 133 155 116 112 112 112 112 112 135 164 144 144 152 152 152 189 1900 5 50 133 133 133 133 155 116 111 111 111 111 111 135 165 144 144 152 152 152 189 190

50 5 50 134 134 134 125 156 117 1 23 45 67 89 136 166 145 145 153 153 153 189 190100 5 50 134 134 134 125 156 117 2 24 46 68 90 136 167 145 145 153 153 153 189 190150 5 50 134 134 134 126 157 118 3 25 47 69 91 137 168 146 146 153 153 153 189 190200 5 50 134 134 134 126 157 118 4 26 48 70 92 137 169 146 146 153 153 153 189 190250 5 50 134 134 134 126 157 118 5 27 49 71 93 137 170 146 146 153 153 153 189 190300 5 50 134 134 134 127 158 119 6 28 50 72 94 138 171 147 147 153 153 153 189 190350 5 50 134 134 134 127 158 119 7 29 51 73 95 138 172 147 147 153 153 153 189 190400 5 50 134 134 134 127 158 119 8 30 52 74 96 138 173 147 147 153 153 153 189 190450 5 50 134 134 134 127 158 119 9 31 53 75 97 138 174 147 147 153 153 153 189 190500 5 50 134 134 134 128 159 120 10 32 54 76 98 139 175 148 148 153 153 153 189 190550 5 50 134 134 134 128 159 120 11 33 55 77 99 139 176 148 148 153 153 153 189 190600 5 50 134 134 134 128 159 120 12 34 56 78 100 139 177 148 148 153 153 153 189 190650 5 50 134 134 134 128 159 120 13 35 57 79 101 139 178 148 148 153 153 153 189 190700 5 50 134 134 134 129 160 121 14 36 58 80 102 140 179 149 149 153 153 153 189 190750 5 50 134 134 134 129 160 121 15 37 59 81 103 140 180 149 149 153 153 153 189 190800 5 50 134 134 134 129 160 121 16 38 60 82 104 140 181 149 149 153 153 153 189 190850 5 50 134 134 134 129 160 121 17 39 61 83 105 140 182 149 149 153 153 153 189 190900 5 50 134 134 134 130 161 122 18 40 62 84 106 141 183 150 150 153 153 153 189 190950 5 50 134 134 134 130 161 122 19 41 63 85 107 141 184 150 150 153 153 153 189 190

1000 5 50 134 134 134 130 161 122 20 42 64 86 108 141 185 150 150 153 153 153 189 1901050 5 50 134 134 134 131 162 123 21 43 65 87 109 142 186 151 151 153 153 153 189 1901100 5 50 134 134 134 131 162 123 22 44 66 88 110 142 187 151 151 153 153 153 189 1901150 5 50 132 132 132 132 163 124 114 114 114 114 114 143 188 151 151 154 154 154 189 1901200 5 50 132 132 132 132 163 124 115 115 115 115 115 143 188 151 151 154 154 154 189 1901250 5 50 132 132 132 132 163 124 115 115 115 115 115 143 188 151 151 154 154 154 189 190




Compositions for Benchmark Model (THERMIX-Konvek)




Results neutronics exercise 1

• Calculations performed with the Diffusion code of ZIRKUS (2D)–Cavities treated via direction dependent diffusion coefficients

• For comparisons: DORT 2D SN-calculations P0- transport corrected–Cavities treated as vacuum

Value ZIRKUS/Diffusion DORT-S16 DORT-S12 DORT-S8 DORT-S6 DORT-S4 DORT-S2k-eff 1.003488 0.992842 0.991904 0.99383 0.997045 0.99376 0.995983Maximum Power density (W/cm3) 12.14 12.2994 12.3250 12.3264 12.3240 12.3330 12.4070Maximum fast flux (n/cm2/s) 2.005E+14 2.041E+14 2.045E+14 2.045E+14 2.043E+14 2.046E+14 2.052E+14Maximum thermal flux (n/cm2/s) 3.055E+14 3.141E+14 3.141E+14 3.153E+14 3.170E+14 3.153E+14 3.171E+14Leakage from core (% per lost neutron) 15.40 15.24 15.28 15.20 15.06 15.20 15.12Leakage from domain (% per lost neutron) 0.209 0.208 0.210 0.211 0.213 0.167




Fast flux density PBMR-400 exercise 1DORT S16

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Bottom




Thermal flux density PBMR-400 exercise 1DORT- S16

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Relative difference DORT S8 to S16 referencesolution (fast group)




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Relative difference DORT S8 to S16 referencesolution (thermal group)




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Relative difference DORT S2 to S16 referencesolution (fast group)




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Relative difference DORT S2 to S16 referencesolution (thermal group)




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fissi

on s

ourc

e di

strib

utio

n, →

TOP

Fission neutron source distribution (DORT) S-16calculation for exercise 1




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istri

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-s16

, →

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Relative difference in fission neutron source distributionbetween S-8 and S-16 calculation




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-s16

, → TOP

Relative difference in fission neutron source distributionbetween S-2 and S-16 calculation




Exercise 2

Heat flow to surface cooling system 974,5 kw! Maximum fuel temperature 996,7 °C!

Inlet helium temperature (°C) 500Outlet helium temperature (°C) 998.2Inlet Pressure (MPa) 9.33944Outlet Pressure (MPa) 9Total pressure drop (kPa) - Inlet to outlet 339.44Pebble bed pressure drop (kPa) - Top / bottom fuel 278Average fuel temperature (°C) 815.28Average moderator temperature (°C) 796.5Average helium temperature in core (°C) 740.61




Pebble surface temperature distributiionexercise 2

-235-1

001203

305407

5096011

60132515

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92,0

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176,

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225

287,

5

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462,

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Surf

ace

Tem

pera

ture

[ °C

]

Radius [m]Heigh

t [m]

900-1000800-900700-800600-700500-600400-500300-400200-300100-2000-100




Flow fieldexercise 2




Remarks exercise 1

• keff strongly dependent from diffusion constants incavity over pebble bed core for diffusion method

• keff about 1% lower for SN (DORT) method,depending from SN order

• SN order from influence of fluxes at outer boundariesand cavities

• For SN calculations the absorber cross sectionsshould be adopted

• transport correction for P0 calculation adequate ?• Treatment of cavities well defined for SN calculations




Remarks exercise 2

• Spatial discretisation could be refined• fuel, moderator and reflector temperatures seem to

be adequate for steady state cases




Remarks exercise 3

• A reproduction of the reference cross section datawas not possible

• Coupled calculations with interpolated cross sectionsled to inconsistent results

• Some interpolated data should be compared

Application and development of tools for HTR neutronics and thermal hydraulics analysis at IKEHTR applicationsCodes availableZIRKUSMain modular ZIRKUS sequence for stationary neutronics - thermal hydraulics calculationsMain modular ZIRKUS sequence for stationary neutronics - thermal hydraulics calculationsZIRKUS-featuresTypical ZIRKUS/THERMIX applicationsZIRKUS optionsZIRKUS modular chainCross section data base and spectral codeMICROX-2 for detailed calculation of spectra in fast, resonance and thermal rangePresent MICROX-2 Library based on JEFF-3.1Thermal neutron scattering on graphiteComparison of measured and calculated spectraTransient calculationsThermal-hydraulic ModelsMacro and Micro model of Heat conductionThermal-hydraulic Modelsused in KIND codes Data Flow to RZKINDFeatures of the KIND CodesExample: Control rod withdrawal(2D RZ-KIND calculation for HTR Modul) Influence of detailed fuel temperature model(control rod velocity: 100 cm/s)Fast transient calculation Further developmentsBenchmark PBMR 400Neutronics/Thermal hydraulicsCompositions for Benchmark Model (THERMIX-Konvek)Results neutronics exercise 1Fast flux density PBMR-400 exercise 1 DORT S16Thermal flux density PBMR-400 exercise 1DORT- S16Exercise 2Pebble surface temperature distributiion exercise 2Flow field exercise 2Remarks exercise 1Remarks exercise 2Remarks exercise 3

w. bernnat, m. buck, n. bensaid, ike, university of stuttgart · w. bernnat, m. buck, n. bensaid,...

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