Presented by Valery Strizhov

MBIR Seminar

Current status of code development for

safety analyses of liquid metal reactors

РОССИЙСКАЯ АКАДЕМИЯ НАУКИнститут проблем безопасного развития атомной энергетики

RUSSIAN ACADEMY OF SCIENCESNuclear Safety Institute (IBRAE)

14-15 November 2019, Dimitrovgrad, RF

Introduction

Federal Target Program “Nuclear Power Technologies of

a New Generation for the years 2010-2015 with a

perspective to 2020”

Goals of program: development of a nuclear power technologies of new generation on the base of fast reactors to increase effectiveness of the uranium and spent fuel use Advanced technologies of fast reactors (LFR & SFR)

New experimental facilities and equipments modernization and development of experimental base

Technologies of advanced fuels fabrication for fast reactors

Materials and technologies of fuel cycle closure for fast reactors

New generation of codes required for the design, construction, operation and safety justification

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Goals of subproject “Codes of new generation”

To provide designers and operating organizations of LMR, developed within the framework of the Federal target program, with the set of new generation codes necessary for design and safety justification Completeness and modern level of description of physical

processes and mathematical algorithms for solving the problem

Modern technology of development and verification of codes, providing quality control of the output product at all stages of the software life cycle

Compliance with the requirements of the Russian regulator (obligatory) and the requirements of foreign rules and regulations (in general)

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•A.A. Bochvar Research

Institute of Inorganic

Materials

•State Scientific Center

«NIIAR»

•All-Russia Research

Institute Of Theoretical

Physics

•All-Russia Research

Institute Of Experimental

Physics

Cooperation

• Rostechnadzor

• Afrikantov Experimental

Design Bureau for Mechanical

Engineering

• Research & Development

Institute of Power

Engineering (NIKIET)

•All-Union Design &

Research Association

•OKB “GIDROPRESS”

• Nuclear Safety Institute

• Institute of Applied

Mathematics

• Institute of Numerical

Mathematics

• National Research

Center IPPE

• Nuclear University

MEPhI

Academy and

univer-sities

ROSATOM

Design Organi-zations

RegulatorROSATOM Industrial Scientific Institutes

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System of Codes of New Generation

Neutronics (ODETTA,

MCU-FR, CORNER,

BPSD, COMPLEX)• Express evaluation

• Engineering calculations

•(diffusion and transport

approaches)

•High-precision

calculations

•(Monte Carlo method,

transport

approximation…)

Fuel and cladding

behaviour code

BERKUT•Fuel rod temperature

distribution

•Microstructure effects in

fuel

•Mechanical deformation

Integral codes:

SOCRAT-BN, EUCLID

Transport module

AEROSOL/LM of activity, fission and

corrosion products in

the coolant circulation

circuits, in the NPP

compartments

Technological processes

included in on-site nuclear

fuel cycle: VIZART

Thermohydraulics

(HYDRA-

IBRAE/LM,

LOGOS, CONV-

3D)•Lumped parameter

codes (1D)

•CFD: RANS, LES, DNS

Impact on the environment

and population: ROM,

ROUZ, SYBILLA

Geomigration: GeRa

Reactor containments:

KUPOL-BR

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Codes development status

Code Comment Status

CRISS 5.3 Code for probabilistic safety analysis Certified

Neutron-physical codes

MCU-FR Neutron-physics code based on the Monte-Carlo method Certification

CORNER Neutron-physics finite differences code based on the kinetic approximation for regular cells Certification

ODETTA Neutron-physics finite elements code based on the kinetic approximation and discrete ordinates for the calculation of reactor shielding

Certification

BPSD Nuclide kinetics code, for calculation of activity and decay heat Certification

COMPLEX Integral code for the radiation safety justification of a FRs installation and fuel cycle facilities Validation

Thermo-hydraulic codes

HYDRA-IBRAE/LM/V1

System thermo-hydraulic code Certified

LOGOS RANS CFD code Certification

CONV-3D DNS CFD code Certified

KUPOL-BR Modeling of the propagation of heat and mass transfer in the system of rooms of NPPs Certification

Fuel performance code

BERKUT Fuel rod code based on the engineering correlations Certified

BERKUT-U Mechanistic fuel performance code Validation

Transport of fission products in environment

ROM Evaluation of the radiation condition during atmospheric transport Certified

ROUZ Code for calculation of the radiation conditions on the industrial site Certification

Sybilla Calculation of irradiation along water pathways Certified

GeRa/V1 Safety validation of the disposal of all types of prepared radioactive wastes Certified

Integral codes

SOCRAT-BN/V1

Comprehensive analysis of SFRs and with oxide fuel for normal operating conditions and design bases accidents without core melting

Certified

SOCRAT-BN/V2 As V1 plus accidents with core degradation of SFRs Certification

EUCLID/V1 Comprehensive analysis of SFR, LFR and LBFR with oxide and (UPu)N fuels for normal operating conditions and design bases accidents without core melting

Certified

EUCLID/V2 As V1 plus accidents with core degradation of SFR and LFR Validation

Codes for modeling processes in closed nuclear fuel cycle

VIZART Code for a calculation of the balance of materials and isotopic flows in a closed nuclear fuel cycle

Validation

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Hierarchy of Neutron Transport Codes

DN3D – 3D multigroup diffusion method

MCU-FR – Monte Carlo method for solving nutron transport equation

ODETTA is a deterministic transport code for radiationshielding problems.

solving 3D steady state multigroup neutron and gamma rays transport equation

finite element method on unstructured tetrahedral meshes

discrete ordinate method

anisotropic scattering can be treated by Pmapproximation

OpenMP technology is applied for parallel computing

CORNER code is designed to calculate the spatial and energy distribution of the neutron flux and its functionals, as well as problems with cavities, large gradients of neutron field and problems with a high degree of attenuation of radiation

Features: solving 3D steady state and

transient multigroup neutron transport equation

different schemes for spatial discretization

SN approximation

Pm Legendre scattering approximation

OpenMP technology is applied for parallel computing

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Hierarchy of TH Models

System level thermo-hydraulic code HYDRA-IBRAE/LM/V1 designed for a computation analysis of non-stationary thermo-hydraulic processes in the loops of nuclear facility) with the subchannel option.

CFD code LOGOS with RANS models of turbulence designed to calculate flows by numerical integration Navier-Stokes system of equations with the method of control volumes.

CFD CONV-3D designed for Direct Numerical Simulation of stationary and non-stationary laminar and turbulent flows of the coolant, as well as heat exchange with solid elements of the equipment.

All codes are designed for simulation of liquid metal coolants (sodium, lead, lead-bismuth), and water

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Modeling of flows in the upper plenum of MONJU facility with CONV-3D Code

OHIRA H., Xu Y., et al. “Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU reactor

vessel” (Proc. of the Int. Conf. on Fast Reactors and Related Fuel Cycles (FR13), Paris, France, March 4–7, 2013),

Sodium temperature at the

points of thermocoupleы

locationVelocity and temperature fields

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Fuel Performance Code

Code BERKUT is designed for multi-scale mechanistic modeling of fuel pins behavior with different types of fuel (UOX, MOX, (UPu)N under normal, transient and accidental conditions

Microscale (fuel grain scale) Evolution of fuel defect microstructure

Gas-filled porosity nucleation and growth

FP generation, radioactive transformations and intragranular transport

Formation of chemical compounds and their distribution over condensed phases

Mesoscale (fuel pellet scale) Intergranular FP transport and release from the pellets

Evolution of the porosity and columnar grain formation

Formation of solid precipitates on the fuel surface

Pellet deformation, cracking and cracks healing

Local fuel melting

Macroscale (fuel rod scale) Heat transfer and exchange with the coolant

Temperature distribution in the fuel, gap and cladding

Gas pressure increase

Changes in rod geometry, deformations and stresses generation

Gap collapse, PCMI and possible cladding failure

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Simulation of BORA-BORA Irradiation Experiment

Two rod types, ‘Pu45’ and ‘Pu60’ were

irradiated

Experimental conditions in BOR-60 reactor

Parameters Pu45 Pu60

Density, %TD 85 85

Pu / (U + Pu), % 42 56,5

Irradiation time, days 900 900

Peak burn-up, at.% 9.4 12.1

Peak linear heat rate,

kW/m

41.9 54.5

Irradiation dose, dpa 43 43

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Environmental code ROM

ROM code is developed for calculation of spreading of radioactive materials in the atmosphere and determining of such parameters as earth concentrations of radionuclides, dose rates after releases of radioactive materials in the gas and aerosol forms into atmosphere The code ROM is based on employing real time

meteorological data

Accounting for 3D time-dependent wind fields

Modeling of dry and wet deposition of aerosols on the surfaces

Performing series of calculation with all possible combinations of the atmospheric parameters (for short-time releases)

Performing multivariant calculations employing site-specific time-dependent series of the meteorological parameters

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ROM code results of analyses

A realistic way to deal with the

situation is to employ real data:

•On the earth relief along the

trace of near surface plume

•time-dependent atmospheric

parameters during the release

period

The map of external dose from

the cloud (mSv) caused by the

accidental release initiated by

the depressurization of SG tube

and failure of isolating fitting

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Environmental code ROUZ

ROUZ has been developed to simulate site airborne and surface contamination fields as a result of gas and aerosol releases

the code takes into account 3D geometry of buildings, atmospheric stability and inhomogeneous turbulence in the atmospheric boundary layer

the code is based on the incompressible Navier-Stokes equations

the code takes into account the stability classes of the atmosphere

the code allows the use of coarse meshes and avoiding mesh refinement in the vicinity of the surface, thereby substantially decreasing the calculation time

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GeRa code development

Code GeRa is an integral code, describing the radioactivity transport in the soils allowingdevelopment of the whole model ranging from the creation of a geological model of the object and ending with the doses calculation for the population due to consumption of water and doses

Long term safety of the deep and surface waste disposals

The correspondence of the proposed solutions to the regulation

geofiltration geomigration

Doses 15

SOCRAT-BN Development

SOCRAT-BN integral code hasbeen developed as multiphysicscode for numerical modeling ofnew generation SFRs (BN-800,BN-1200) behavior under DBAand BDBA conditions from initialset of events up to release of FPto environment

0 4 8 12 16 20 24 28 32 36 40 44 48 52

-4

-2

0

2

4

6

8

10

12

14

Fuel melting

Steel melting

Start Boiling

Pow

er, re

lati

ve

unit

Time, s

Power

Power

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EUCLID/V1 code development

ЕВКЛИД - integrated multiphysics code for numericalsimulation of the behavior of newgeneration reactors with liquidmetal coolants (BN-800, BREST,BR-1200) in normal operation,design and beyond designaccidents

Main components: HYDRA-IBRAE/LM/V1 (sodium,

lead, lead-bismuth, water)

BERKUT (UOX, MOX, (UPu)N)

Neutron physics (DN3D,CORNER)

Burn up BPSD

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BN-600 Scram Simulation

Summary

Within the framework of the sub-project “Codes of the New Generation” of “Breakthrough” project the first stage of the development of codes system is underway, which allows for solving different safety problems

Standalone high computational codes have been developed as well as integral codes, which are built through the integration of individual components of a system of codes into a single software package

The system of codes allows the conducting consistent multi-physical and multi-scale analysis of operating and emergency regimes, including conducting PSA levels 1, 2 and 3.

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