user guide - 西安交通大学sarax.xjtu.edu.cn/user_guide.pdf · n nodal method using...

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SARAX-2.0e

System for Advanced Reactor Analysis at Xi’an Jiaotong University

User Guide

by

Youqi Zheng

Xianan Du Zi’an Zhai Linfang Wei

Xi’an Jiaotong University

2018.4.8

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PREFACE

The SARAX (System for Advance Reactor Analysis at Xi’an Jiaotong University) code system is a new code system developed to model and simulate the neutronics behaviors of reactors and subcritical facilities with fast spectrum. It consists of 4 parts including TULIP for cross-section generation, LAVENDER for steady state core analysis, DAISY for transient analysis and COLEUS for S&U analysis. The code system is capable of doing the neutronics analysis for SFRs, LFRs, LBE cooled ADS reactors and part of GFRs.

This documents is written to introduce how to use the free released version of SARAX code system, i.e. SARAX-2.0e. It consists of the introduction about how to install the code package, which modules are in this package and how to model a SFR using these modules. The description of input cards together with a sample for modeling a 1000MW SFR from the cross-section generation to the core calculation in steady state are supplied.

The code released together with this document are the achievements of the members’ years of effort in the NECP lab. in Xi’an Jiaotong University. We appreciate their great contributions. And, special thanks to Prof. Won Sik Yang from Purdue University and Prof. Deokjung Lee from UNIST who have given great advice and comments to the method study and code development.

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Contents

1. Introduction ........................................................................................................... 1 2. Contents in the released file ................................................................................. 3 3. Requirement and Installation .............................................................................. 4

3.1 Requirement ..................................................................................................... 4 3.2 Installation........................................................................................................ 4 3.3 Running the code system ................................................................................. 6 3.4 Uninstallation ................................................................................................... 7

4. Main modules in SARAX-2.0e ............................................................................. 8 4.1 Nuclear Data Library ....................................................................................... 8 4.2 Depletion information ...................................................................................... 8 4.3 Cross-section generation .................................................................................. 9 4.4 Core analysis in steady state ............................................................................ 9 4.5 Interfaces ........................................................................................................ 10

5. Modeling a Sodium cooled fast reactor ............................................................. 11 5.1 Homogeneous model for homogenization ..................................................... 11 5.2 Heterogeneous model for homogenization .................................................... 11 5.3 Core geometry specification .......................................................................... 12 5.4 Reactivity calculation..................................................................................... 13

6. Input description ................................................................................................. 14 6.1 Contents in input and way to describe ........................................................... 14 6.2 Input table for TULIP .................................................................................... 16 6.3 Input table for LAVENDER ........................................................................... 22

7. Sample input ........................................................................................................ 31 7.1 Specification of the calculated problem ......................................................... 31 7.2 Sample TULIP input card .............................................................................. 32 7.3 Sample LAVENDER input card .................................................................... 34

8. Sample output...................................................................................................... 37 8.1 Sample TULIP output .................................................................................... 37 8.2 Sample LAVENDER output .......................................................................... 39

9. Reference ............................................................................................................. 41 Appendix A List of Isotope Symbol in TULIP code ................................................ 42 Appendix B Interface to SN code using ANISN format .......................................... 46 Appendix C Interface to Use Hybrid Method ......................................................... 47

User Guide of SARAX-2.0e Code System

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1. Introduction

The SARAX (System for Advance Reactor Analysis at Xi’an Jiaotong University) code system is a code system developed for the neutronics analysis of reactors and external source driven subcritical facilities with fast spectrum. It is based on the state-of-the-art computational methods and aimed to service as both the industrial design tool and research tool for institutes and universities.

The SARAX, of which the development started in 2012 with the foundation of National Natural Science Foundation of China, was original designed for the LBE cooled ADS reactors. The version SARAX-1.0 was firstly published in the domestic reactor physics conference (CORPHY2014) in 2014, which invoked an external Monte-Carlo code to generate the homogenized cross-section.

A new cross-section generation part was completed in 2016, and the code system was extended to be available for SFR, LFR and LBE ADS reactors in the same year. In 2017, the part for sensitivity and uncertainty (S&U) analysis was developed and a new library SARALIB with adjusted nuclear data for MOX-fueled sodium fast reactor was completed for the purpose of industrial application.

Now, four main parts constitute the SARAX code system, including:

- TULIP: the part for cross section generation. The ultrafine group method was adopted for the education and industrial versions. In the latter one, finer models for leakage and reactivity conservation were developed. A hybrid method was proposed and merged in the research version.

- LAVENDER: the part for core analysis in steady states, which is capable of doing keff and power distribution calculation, depletion calculation, reactivity evaluation and fuel management etc. The SN nodal method using triangular-z mesh was used for the flux solver for all versions. The parallel capability and new solver using hexagonal-z mesh are under testing for the industrial version.

- DAISY: the part for transient analysis, which integrates transient single-channel T-H modules for sodium, lead and LBE coolant. The point kinetics method was used for the education and industrial versions with differences of dealing with the reactivity feedback. The predict-correction quasi-steady method was used for the research version.

- COLEUS: the part for S&U analysis. The calculation was based on the generalized perturbation theory. The uncertainty of keff, power distribution and reactivity uncertainties can be quantified. The similarity analysis and nuclear data adjustment functions have been completed in the industrial version.

The newest version is SARAX-2.0, which is a commercial version for industrial


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application. The system has full functions for neutronics design works and is matched with a special nuclear data library named SARALIB.

Besides the commercial version, a free of charge version, as SARAX-2.0e, has been simplified from the complete version, which aims to be able to perform basic calculations for SFR. This version, SARAX-2.0e, has been released together with this document and can be downloaded from the website: http://sarax.xjtu.edu.cn for free.

This work was carried out partially under the financial support of the National Nature Science Foundation of China (Approved number 111051044, 91126005, 11475134, and 11775170). Six doctors contributed a lot to the code development in the past 5 years, who are:

Dr. Xianan Du, who worked on the module of cross-section generation and will defense the Doctoral degree in Jun. 2018 Dr. Chenghui Wan, who worked on the adjusted nuclear data library and defensed the doctoral degree in Mar. 2018 Dr. Yong Liu, who worked on the S&U analysis module and defensed the doctoral degree in Sep. 2017 Dr. Shengcheng Zhou, who worked on the steady state core analysis module and defensed the doctoral degree in Mar. 2017 Dr. Yunlong Xiao, who also worked on the steady state core analysis module and defensed the doctoral degree in Mar. 2017 Dr. Mingtao He, who worked on the transient analysis module and defensed the doctoral degree in Mar. 2017

Seven new members have joined the development team since 2016, who are working on the method/model improvement and code maintaining. New methods and more modern programing techniques are being tested and will come out in the near future for better accuracy, easier use and wider feasibility.


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2. Contents in the released file The released code package consists of documents, executable codes and sample exercises, which include:

1. User Guide of SARAX-2.0e

2. Copyright Announcement

3. References of the methods and V&V works used in SARAX

4. Executable codes for Windows OS (32 bit) with library packaged in

5. Executable codes for Windows OS (64 bit) with library packaged in

6. Executable codes for Mac OS X Platform (64 bit) with library packaged in

7. Samples input and output files of OECD SFR benchmark

Following figure illustrates the contents in the released USB flash memory disk:


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3. Requirement and Installation

3.1 Requirement

SARAX-2.0e includes binary executable codes for Mac OS X, Windows XP+.

Minimum requirements

The installation of SARAX-2.0e code system requires at least 4 GB RAM and 10 GB disk space. The main frequency of CPU needs to be higher than 2.0 GHz. Also, additional disk space is required to store output results when running SARAX-2.0e code.

Recommended requirements

To achieve better performance of running SARAX-2.0e code, 8 GB or more RAM is recommended. Multi-core CPU, which the main frequency is higher than 3.0 GHz, is also recommended to reduce time consumption. The requirement of hard disk space is as same as Minimum requirements.

It should be noticed that executable codes can run with Windows OS under 32-bit architecture. However, there is memory limitation of allocated array. Therefore, it is not suitable to use 32-bit executable code for large problem.

The executable codes for all platforms were created using INTEL IFORT 13.0.

3.2 Installation

1）Select specific installation package, double click;

2）Select installation language, press OK;

3）Welcome Page, press Next;


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4）Choose an arbitrary directory to unpack the file on Install Directory Page, then press Next;

5）Installing Page: 5 minutes are needed. (Depends on computer);


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6) Finish Page.

The Installation Directory includes files as follows:

bin\ Executable files

library\ Nuclear data required by SARAX-2.0e code

uninstall.exe Uninstall execution

3.3 Running the code system

When the installation is finished, the path of executable codes and nuclear data libraries will be added into environment variable automatically as follows:

For executable codes, PATH is added. For library, SARAX_LIBRARY is created.

For Windows OS, the executable codes are tulip.exe and lavender.exe, respectively. For Mac OS, the executable codes are tulip and lavender. Since the relative environment variables are already added, each code can run anywhere by using command line.

The way to run example OECD MOX1000 SFR benchmark using command line is as follows:

cd example/OECD_SFR_benchmark/MOX1000 tulip or tulip.exe


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lavender or lavender.exe

Besides, you can copy the executable codes together with the input files and double click the executable codes.

3.4 Uninstallation

If you want to uninstall SARAX-2.0e code, you can double click uninstall.exe to perform uninstallation.

To confirm the uninstallation, press Y.

Uninstalling Page

When Uninstallation is finished, relative environment variables will be deleted.


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4. Main modules in SARAX-2.0e

4.1 Nuclear Data Library

The nuclear data library in SARAX-2.0e was generated by using NJOY2016 based on the ENDF/B-VII.0 evaluated nuclear data files. In this version, only cross-sections of neutron are included. Totally, 351 isotopes are produced, which is listed in the Appendix A.

A special post-processing code has been developed to generate the library for SARAX. For each nuclide, 5 files are generated:

.pw Point-wise cross sections .em Elastic scattering function .nes 1968-group inelastic scattering matrices, P1 order .ug 1968-group total cross section, .kp 1968-group delayed neutron data

Besides, the 1968-group smooth cross-section data for all heavy isotopes are stored in one single file, named nuclide_info.

4.2 Depletion information

11 heavy isotopes are considered in depletion chain, include 235U, 236U, 238U, 237Np, 239Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu,241Am. 166 fission products are lumped for each heavy isotope. Therefore, 11 lumped fission products are used in depletion calculation. In the lumped fission product, only solid fission products are considered. The fission yield data of each fission product are selected from JNDC-V2.0 library.

Following figure illustrates burn-up chain used in the LAVENDER code:

zmA ,n γ z

m+1

z+1m+1

z+1m

β −

zm-1

, 2n n

FP

fσ

z-2m-4

α

U235 ,n γ

U236 ,n γ

Np237 ,n γ

Pu238 ,n γ

Pu239 ,n γ

Pu240 ,n γ

Pu241 ,n γ

Pu242

Am241

β −

U238 ,n γ

Np239 ,n γ

β −


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4.3 Cross-section generation

In the code package, TULIP is developed for the cross-section generation. TULIP code allows users to specify problem materials in a homogeneous mixture or in a one-dimensional (1D) cylinder/slab geometry. For the homogeneous mixture, the eigenvalue calculation, the fixed-source calculation and the critical buckling search calculation could be done. For the 1D problem, only eigenvalue calculation could be done.

For each problem, 1968 ultrafine-group self-shielded cross sections are generated based on ultrafine-group method. The point-wise libraries are directly used when performing self-shielding calculation. To obtain detailed neutron spectrum in a homogeneous mixture, the narrow resonance (NR) approximation is used. The escape cross section is used to take the heterogeneity effect into account. The elastic scattering transfer matrices are calculated using self-shielded group-averaged elastic scattering cross section and elastic scattering function. The consistent P1 order N extended transport equations are solved to determine 1968-group neutron spectrum in homogeneous mixture problem. The collision probabilities method is used to obtain neutron spectrum for 1D cylinder/slab problem.

After obtaining neutron spectrum, the cross sections would be homogenized into few groups and/or few regions. The homogenization procedure is based on the conservation of reaction rate. Currently, 33 energy group structures is used for few-group cross sections generation.

4.4 Core analysis in steady state

In the code package, LAVENDER is developed for the core calculation in steady state. The three-dimensional (3D) steady state core analysis is performed with LAVENDER code. The reactor that consists of hexagonal assembly or rectangle assembly could be simulated. The whole-core, 1/2 core, 1/3 core, 1/4 core, or 1/6 core could be built by User’s definition. For the Education version, LAVENDER code can only be used in the analysis of eigenvalue problems in FR.

In LAVENDER code, the SN nodal transport method in triangular-Z geometry is adopted. The hexagonal-based CMFD method and OpenMP parallel calculation in angular sweeping are used to reduce calculation time consumption.

LAVENDER code also performs depletion analysis to investigate the reactivity swing during the lifetime. During the depletion analysis, the sub-step method and predict-correct method are alternatively used. However, the depletion calculation is only for specific core arrangement. The fuel arrangement problems, such as the refueling problem, could not be simulated.

Two methods, including the direct method and the perturbation method, are implemented to calculate reactivity changes. When the reactivity change of different core state is needed, the adjoint calculation of 3D core can be alternatively performed


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with LAVENDER code. In the current version, only the coolant voided reactivity worth can be calculated by using exact perturbation theory. The other reactivity coefficient could be obtained by perform two steady state core calculations.

4.5 Interfaces

Interfaces have been developed for the extended calculations. Two examples are given in the Appendix B and C:

1) Interface to SN code. TULIP can output cross sections in ANISN format so that the code, like ANISN and DORT, can directly read these cross sections. It is usually used for finer consideration of leakage/interaction between neighbor material zones while the group condensation or homogenization is done. In SARAX-2.0, a new 2D/3D SN solver, Hydra, has been integrated but not included in the free released SARAX-2.0e.

2) Interface to Monte-Carlo code OpenMC. In the free released SARAX-2.0e, the TULIP module has been simplified. However, we supply an interface to apply the new hybrid method to generate better cross-sections, which combines the continuous energy calculation and the traditional ultra-fine method.


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5. Modeling a Sodium cooled fast reactor

5.1 Homogeneous model for homogenization

Since the mean free path of fast neutron is longer than that of thermal neutron, the homogeneous assembly model is usually used in generating the homogenized cross-sections. A homogeneous assembly model can be built by define the geometry type in the input file,

geometry_type homo

Then, the composition of this homogeneous mixture needs to be defined. Since an assembly always consists of different materials, the nuclide density of each isotope should be homogenized by volume-weighted.

For example, a homogeneous mixture consists of 3 isotopes, including 235U, 238U, and 23Na. The nuclide densities are given as follows:

235U 238U 23Na

9.10061E-06 6.41086E-03 7.40989E-03

The input card will be written as:

!=================================================================================

PROBLEM:

number 1

! this part is for case name

CASENAME:

case1

CONTROL:

material_num 1

geometry_type homo

MATERIAL:

1 3

U235 9.10061E-06 1300.

U238 6.41086E-03 1300.

Na23 7.40989E-03 700.

!=====================================(END:)======================================

END:

5.2 Heterogeneous model for homogenization

If the heterogeneity cannot be ignored, the heterogeneous model should be used. Currently, the TULIP code can be only used for 1D problem, including 1D-cylinder and 1D-slab problems. Here, we show an example to make the equivalent model for a heterogeneous subassembly in the following figure. The equivalent radius should be


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calculated by preserving the area of each material like fuel, cladding and coolant. The nuclide number density should be given accordingly.

Fuel Cladding & Wrapper tube Coolant

To define a heterogeneous assembly, first we should select the geometry type:

geometry_type 1D-cylinder or 1D-slab

Then total number of material should be provided:

material_num a integer parameter

After define each material, the parameters for geometry should be defined. The figure shows two 3-region problems.

0 r0.4 1.0 1.6

1 2 1

0 x0.4 1.0 1.6

1 2 1

Therefore, the definition for each parameter should be given as:

total_region 3 region_material 1 2 1 region_size 0.4 0.6 0.6 thickness of each region

5.3 Core geometry specification

The core configuration is mainly divided into two parts, the axial material distribution of each assembly, and the radial location of each assembly. To define the axial core configuration, the assembly needs to be divided into a number of layers in the axial direction firstly:

! n_layer height (bottom-->top) layer38 2*17.8800 8*14.04785 10*11.494 10*11.997 4*13.1100 4*11.1750


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Then, the material for each axial mesh is determined by using number index of material: FA_type 1 2*16 8*17 2*1 2*2 2*3 2*4 2*5 14*19 4*18 FA_type 2 2*16 8*17 2*6 2*7 2*8 2*9 2*10 14*19 4*18 FA_type 3 2*16 8*17 2*11 2*12 2*13 2*14 2*15 14*19 4*18 FA_type 4 2*16 8*17 24*20 4*18 FA_type 5 2*16 8*17 24*21 4*18 FA_type 6 2*16 8*17 10*23 10*22 8*23

Finally, the layout of radial core is listed by using number index of assembly: assign_FA

6 1 1 6 2 2 6 3 4 4 5 1 1 1 1 2 2 2 3 4 4 5 1 1 1 1 2 2 2 3 3 4 4 5 6 1 1 6 2 2 6 2 3 4 4 5 2 2 2 2 2 2 2 2 3 4 4 5 2 2 2 2 2 2 2 3 3 4 4 5 6 2 2 6 2 2 6 3 3 4 4 5 3 3 3 2 2 3 3 3 4 4 4 5 4 4 3 3 3 3 3 4 4 4 4 5 4 4 4 4 4 4 4 4 4 4 5 5 5 5 4 4 4 4 4 4 4 5 5

5 5 5 5 5 5 5 5

5.4 Reactivity calculation

In SARAX-2.0e, two ways to calculate the reactivity coefficients are supplied.

i. Direct way

To evaluate the reactivity changes due to the change of core state, two different core calculations will be performed. To get the results, two sets of self-shielded cross sections are generated first. One is the standard case and the other is perturbed case. Then the core calculation is done twice by using different set of cross sections. In case of control rod worth calculation, one set of cross sections can be used.

ii. Perturbation way

This way is to get the reactivity using the perturbation theory. For example, if the exact perturbation theory is applied in calculating the sodium void reactivity, only the position where the voided region locates and how much the coolant fraction loses should be specified in the input card of LAVENDER without running twice TULIP for non-voided and voided cases.


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6. Input description

6.1 Contents in input and way to describe

The input file of TULIP code is named as tulip.inp.

The input file of LAVENDER code is named as lavender.inp.

The name of input file should not be changed.

Each input file will be divided into several blocks.

TULIP code contains 6 blocks as follows:

PROBLEM: Definition of the problem, such as the number of case, group, et al; CASENAME: Name of current case, a directory will be created using it; CONTROL: Logical definition of current case; MATERIAL: Material description of current case; GEOMETRY: Geometry description of current case; END: A symbol to end current case;

LAVENDER code contains 8 blocks as follows:

CASENAME: Name of current case; CONTROL: Logical definition of current case; METHOD: Description of method using in transport solver; MATERIAL: Define the cross sections for each material; GEOMETRY: Define the core geometry and material distribution; DEPLETION: Define the calculation condition of depletion calculation; REACTIVITY: Define the calculation condition of reactivity calculation; END: A symbol to end current case;

The way to describe the input information is based on [keyword + attribute]. In each block, there are several keywords can be defined.

Following table gives the description of character used in input card and the explanation of input card presented in next sections.

Input Deck Format

Character Description

! Comment Line

CS Character String

Es Variable using Scientific Notation

I, N Integer Variable

L Logical Variable (T/F)


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R Real Variable

W Nuclide with Mass Number

To describe the input table for TULIP and LAVENDER code, some blocks and keywords are required during writing an input card. When there is no required symbol, it means this block or keyword could be ignored when writing an input card. When the keywords are ignored, the default value will be used in the following calculations.


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6.2 Input table for TULIP

PROBLEM Block

Title Input Parameters Contents Default Description

PROBLEM:

required

start of general card

PROBLEM:

number

required

I - (None) total case number

number 23

mode 2D or 3D 3D subsequent calculation method,

2D::used for 2D core calculation

3D::used for 3D core calculation

mode 3D

library Cs default library path

Library F:\tulip_data\

group_info I, I, I 1968,33,1 ultrafine group number, few group

number, order of PN

group_info 1968 33 1

tr_correction L True transport correction switch

tr_correction T

output_xs L,L False, False, decimal format switch, ANISN

format switch

output_xs T F

multi_temperature L, N, I1... IN False, -, -... multi-temperature switch,

temperature point number,

temperature point level

multi_temperature F 3 600. 900. 1200.


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CASENAME Block


CASENAME:

required

Cs Default_case title of single card

CASENAME:

mat1


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CONTROL Block


CONTROL:

required

CONTROL:

condensation L, L True, True group condensation switch, region

condensation switch

condensation T T

material_num

required

I - total number of materials

material_num 3

geometry_type

required

homo/1D-slab/

1D-cylinder

homo geometry type of the case

geometry_type 1D-cylinder

calculation_mode eigen/fixed/

buck

eigen eigenvalue or fixed source or

bucking search problem

calculation_mode eigen

external_source Cs leakage.out* valid if calculation_mode is set as

fixed

external_source leakage.out

external_flux L, I, Cs False, -, - external flux input switch, flux

index, file path

external_flux F 1 zoneflux.txt


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MATERIAL Block


MATERIAL:

required

MATERIAL:

- I, I -, - material ID, total nuclide number

- W, Es, R -, -, - nuclide name, nuclide density,

temperature(unit: K)

1 9

O16 4.24384E-02 523.

U235 4.05533E-03 523.

U238 1.34125E-02 523.

Pu238 3.73868E-06 523.

Pu239 2.86038E-03 523.

Pu240 7.12945E-04 523.

Pu241 9.82312E-05 523.

Pu242 2.02221E-05 523.

Am241 2.36060E-05 523.


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GEOMETRY Block


GEOMETRY: GEOMETRY:

total_region N - total number of geometry regions

total_region 8

region_material I1... IN -... material ID of each geometry region

region_material 1 2 3 4 1 2 3 4

region_size R1... RN -... x or r thickness of each geometry

region(unit: cm)

region_size 0.28109 0.03514 0.16141 0.04537

0.34163 0.04537 0.26145 0.04538

region_condense I, N*I -... condense the last n regions to one

region

region_condense 1 8*1

region_mesh N*I N*1 number of meshes for each region

region_mesh 8*2


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END Block


END

required

end of case

END:


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6.3 Input table for LAVENDER

CASENAME Block

Card Field Default Description

CASENAME:

required

INP_Title NULL Title

CASENAME:. MOX-1000 Benchmark Problem


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CONTROL Block


STEADY

required

Rated_power None Rated power of problem (unit:W)

Power_frac None Power fraction in percentage (unit:%)

STEADY 1.0E8 80

DEPLETION

Is_depletion F Do depletion calculation or not

DEPLETION F

REACTIVITY

Is_reactivity F Do reactivity calculation or not

REACTIVITY F


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METHOD Block


QUADRATURE

required

Is_60degree 1 1: level symmetrical quadrature

2: 60 degree symmetrical quadrature

sn 4 Order of SN

theta None Calculation degree when choose 60 degree symmetrical

quadrature

QUADRATURE 2 4 120

STEADY

Is_LW F Apply LW acceleration method or not

STEADY F

ERROR_TYPE Error_type 2 1: 1-norm

2: 2-norm

3: infinite norm

ERROR_TYPE 3

ERROR_EIGEN

required

Max_inner 10 Maximum cycles of inner iteration

Max_outer 100 Maximum cycles of outer iteration

Error_inner_flux 5.0E-6 Maximum relative error limit of flux in the inner iteration

Error_outer_flux 1.0E-6 Maximum relative error limit of flux in the outer iteration

Error_eigen 1.0E-5 Relative error limit of eigenvalue

ERROR_EIGEN 13 500 5.0E-6 1.0E-6 1.0E-5

OPENMP nthread 4 Thread number used in OPENMP calculation

OPENMP 4

CMFD Cmfd_mg T Perform CMFD acceleration for eigenvalue problem

NOTE: only for hexagonal geometry

CMFD T


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MATERIAL Block


MAT_NUM

required

Mat_num None Number of materials

MAT_NUM 11

ENERGY_GROUP

required

ng None Number of energy groups

Scat_order 0 Order of scatter cross sections

ENERGY_GROUP 33 0


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GEOMETRY Block


TYPE

required

Mesh_type None 1: rectangular arrangement

2: hexagonal arrangement

TYPE 2

LAYER

required

Layer_num 1 Number of layers

Height (layer_num) Height of each layer

LAYER 10 2*7 3*5 4*6 6*4

REF_AXIAL

required

Layer_bottom 0 Number of layers reflectors filled in at the bottom of core

Layer_top 0 Number of layers reflectors filled in at the top of core

REF_AXIAL 5 7

BC_AXIAL

required

Axial(1) None Bottom boundary condition

Axial(2) None Top boundary condition

NOTE: 0: vacuum

1: reflective

BC_AXIAL 0 0

HEX_DIM

required

degree 60 Range of calculation in degree

NOTE: 60/90/120/180/360

ring None Rings of core

NOTE: include the central assembly

pitch None Pitch of assembly (cm)

HEX_DIM 120 12 16.2471

HEX_CONF

required

Position None Order of each line left-most assembly in the hexagonal

arrangement determined in the HEX_DIM

Conf None Array of each line assemblies from left to right

NOTE: a positive number represents this position should

be

filled in an assembly

HEX_CONF 1 6 6 6 6 2 2 6 3 4 4 5

1 6 6 6 6 2 2 2 3 4 4 5

1 6 6 6 6 2 2 2 3 3 4 4 5

1 6 6 6 6 2 2 6 2 3 4 4 5

1 2 2 2 2 2 2 2 2 3 4 4 5


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1 2 2 2 2 2 2 2 3 3 4 4 5

1 6 2 2 6 2 2 6 3 3 4 4 5

1 3 3 3 2 2 3 3 3 4 4 4 5

1 4 4 3 3 3 3 3 4 4 4 4 5

1 4 4 4 4 4 4 4 4 4 4 5 5

1 5 5 4 4 4 4 4 4 4 5 5

3 5 5 5 5 5 5 5 5

HEX_MESH

required

Mesh_size 1 1: one hexagon is divided into six triangles

2: one hexagon is divided into ten triangles

HEX_MESH 2

HEX_BC

required

Bc_symline None Boundary condition of symmetric surface

0: vacuum ; 1: reflective ; 2: rotational

Bc_outer None Boundary condition of outer surfaces

0: vacuum ; 1: reflective ;

HEX_BC 1 0

FA_TYPE

required

N_type None ID of each type assembly

Axial_config(layer_num) None Axial material arrangement from the bottom to the

top

FA_type 1 2*16 8*17 2*1 2*2 2*3 2*4 2*5 14*19 4*18

FA_type 2 2*16 8*17 2*6 2*7 2*8 2*9 2*10 14*19 4*18

FA_type 3 2*16 8*17 2*11 2*12 2*13 2*14 2*15 14*19 4*18

FA_type 4 2*16 8*17 24*20 4*18

FA_type 5 2*16 8*17 24*21 4*18

FA_type 6 2*16 8*17 10*23 10*22 *23

ASSIGN_FA

required

Assign None radial arrangement of different types assemblies

assign_FA 6 1 1 6 2 2 6 3 4 4 5

1 1 1 1 2 2 2 3 4 4 5

1 1 1 1 2 2 2 3 3 4 4 5

6 1 1 6 2 2 6 2 3 4 4 5

2 2 2 2 2 2 2 2 3 4 4 5

2 2 2 2 2 2 2 3 3 4 4 5

6 2 2 6 2 2 6 3 3 4 4 5

3 3 3 2 2 3 3 3 4 4 4 5

4 4 3 3 3 3 3 4 4 4 4 5

4 4 4 4 4 4 4 4 4 4 5 5

5 5 4 4 4 4 4 4 4 5 5

5 5 5 5 5 5 5 5


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DEPLETION Block


SCHEME

Method 1 Scheme of burnup calculation

1: substep 2: predict-corrected

SCHEME1

POWER Step_num None Number of burnup steps

POWER 11

TIME Time(step_num) None Time of each burnup step (unit:day)

TIME 11*32.5

HISTORY Relative_pow(step_num) None Relative power of each burnup step (unit:%)

HISTORY 11*100


29

REACTIVITY Block


VOID_POSITI

ON

Method None Sign the void positions in the core

NOTE: a positive number represents this position

appear coolant void.

VOID_POSITION 0 1 1 0 2 2 0 3 0 0 0

1 1 1 1 2 2 2 3 0 0 0

1 1 1 1 2 2 2 3 3 0 0 0

0 1 1 0 2 2 0 2 3 0 0 0

2 2 2 2 2 2 2 2 3 0 0 0

2 2 2 2 2 2 2 3 3 0 0 0

0 2 2 0 2 2 0 3 3 0 0 0

3 3 3 2 2 3 3 3 0 0 0 0

0 0 3 3 3 3 3 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0

VOID Void fraction None Void fraction (unit: %)

Nuclide number None Number of nuclides in the coolant

Nuclide name None Names of all nuclides in the coolant

VOID 86 1 Na23


30

END Block


END

required

NONE None The keyword of the end of input file

END:


31

7. Sample input

7.1 Specification of the calculated problem

This problem specifies a MOX-fueled 1000MW SFR (MOX-100) referred from the OECD SFR benchmark. The 1/3 core radial layout and axial material distribution are shown in Fig. 7.1. Considering the axial material distribution, there are 23 materials in total in the MOX-1000 core. The definition of each case in the sample input file and its corresponding material are list in the Material Definition table. In the calculation, the homogeneous assembly model is used. The definition of depletion calculation is 328.5 days with full nominal power rating.

Inner Core

Middle Core

Outer Core

Reflector

Primary Control

Secondary Control

Shield

Lower

Structure

LowerReflector

Fuel 1

Fuel 2

Fuel 3

Fuel 4Fuel 5

GasPlenum

UpperStructure

Fig. 7.1 Core profile of MOX-1000 test case

Table 7.1 Material Definition

Real Core Case1~Case5 Fuel zones of inner core assembly from bottom to top Case6~Case10 Fuel zones of middle core assembly from bottom to top Case11~Case15 Fuel zones of outer core assembly from bottom to top

Case16 Lower structure Case17 Lower reflector Case18 Upper structure Case19 Upper gas plenum Case20 Radial reflector Case21 Shield Case22 Absorber Case23 Empty duct


32

7.2 Sample TULIP input card

!=================================================================================

! total case number

PROBLEM:

! Num

number 23

!=================================================================================

! case 1

!=================================================================================


CASENAME:

case1

!=================================================================================

! --this part is global case control logical

CONTROL:

! Number

material_num 1

! Geometry

geometry_type homo

!=================================================================================

! --this part is material description

MATERIAL:

! ID nuclide_number

1 42

! nuclide_name nuclide_density temperature

U234 6.75807E-07 1300.

U235 9.10061E-06 1300.

U236 1.12669E-06 1300.

U238 6.41086E-03 1300.

Np237 1.16922E-05 1300.

Pu236 1.10068E-10 1300.

Pu238 4.30541E-05 1300.

Pu239 7.38223E-04 1300.

Pu240 5.20158E-04 1300.

Pu241 8.34209E-05 1300.

Pu242 1.12965E-04 1300.

Am241 4.17351E-05 1300.

Am242m 3.51718E-06 1300.

Am243 3.73015E-05 1300.

Cm242 2.14971E-06 1300.

Cm243 2.56213E-07 1300.

Cm244 2.82305E-05 1300.

Cm245 8.29936E-06 1300.


33

Cm246 5.02407E-06 1300.

O16 1.69558E-02 700.

Na23 7.40989E-03 700.

Fe54 1.04032E-03 700.

Fe56 1.63842E-02 700.

Fe57 3.91461E-04 700.

Fe58 5.89873E-05 700.

Cr50 1.14553E-04 700.

Cr52 2.22621E-03 700.

Cr53 2.53824E-04 700.

Cr54 6.32566E-05 700.

Ni58 7.52455E-05 700.

Ni60 2.87596E-05 700.

Ni61 1.24979E-06 700.

Ni62 3.95988E-06 700.

Ni64 9.96307E-07 700.

Mo92 8.49589E-05 700.

Mo94 4.84866E-05 700.

Mo95 8.43153E-05 700.

Mo96 8.86598E-05 700.

Mo97 5.07393E-05 700.

Mo98 1.27546E-04 700.

Mo100 5.16512E-05 700.

Mn55 1.17741E-04 700.

!=====================================(END:)======================================

END:


34

7.3 Sample LAVENDER input card !===============================================================================


CASENAME: "OECD_MOX-1000"

!===============================================================================

! --this part is global case control logical

CONTROL:

! power percent

steady 3.33E8 100.0

! is_depletion

depletion T

!===============================================================================

! --this part is calculation method

METHOD:

! type SN sector

quadrature 2 4 120

! is_LW

steady F

! max_inner max_outer rms_inner rms_outer rms_fission_rate rms_eigen

error_eigen 3 500 5.0E-6 1.0E-5 1.0E-5 1.0E-5

! nthread

openmp 8

! cmfd_mg

CMFD T

!===============================================================================

! --this part is material parameter

MATERIAL:

! num is_external is_marco

mat_num 23 T F

! ng scat_order

energy_group 33 0

!===============================================================================

! --this part is core geometry

GEOMETRY:

! type(0-ansys/ 1-rec/ 2-hex)

type 2

! n_layer hz (bottom-->top)

layer 38 2*17.8800 8*14.04785 10*11.494 10*11.997 4*13.1100 4*11.1750

! boundary%bottom boundary%top

bc_axial 0 0

ref_axial 10 18

! degree ring pitch

hex_dim 120 12 16.2471


35

! FA_conf (startswith xxx)

hex_conf 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1

3 1 1 1 1 1 1 1 1

! meshsize (1/ 2)

hex_mesh 1

! radial bc (inner/ outer) 0-vaccum; 1-reflector; 2-symmetry;

hex_bc 1 0

FA_type 1 2*16 8*17 2*1 2*2 2*3 2*4 2*5 14*19 4*18

FA_type 2 2*16 8*17 2*6 2*7 2*8 2*9 2*10 14*19 4*18

FA_type 3 2*16 8*17 2*11 2*12 2*13 2*14 2*15 14*19 4*18

FA_type 4 2*16 8*17 24*20 4*18

FA_type 5 2*16 8*17 24*21 4*18

FA_type 6 2*16 8*17 10*23 10*22 8*23

assign_FA 6 1 1 6 2 2 6 3 4 4 5

1 1 1 1 2 2 2 3 4 4 5

1 1 1 1 2 2 2 3 3 4 4 5

6 1 1 6 2 2 6 2 3 4 4 5

2 2 2 2 2 2 2 2 3 4 4 5

2 2 2 2 2 2 2 3 3 4 4 5

6 2 2 6 2 2 6 3 3 4 4 5

3 3 3 2 2 3 3 3 4 4 4 5

4 4 3 3 3 3 3 4 4 4 4 5

4 4 4 4 4 4 4 4 4 4 5 5

5 5 4 4 4 4 4 4 4 5 5

5 5 5 5 5 5 5 5

!===============================================================================

! --reserve micorscopic-depletion

DEPLETION:

! method (1=substep 2=PC)

scheme 1

! Step Num

power 10

! Time of each step

time 10*32.5


36

! Relative Power of each step

history 10*100

!=====================================(END)=====================================

END:


37

8. Sample output

8.1 Sample TULIP output

When TULIP process is finish, two directories will be created automatically, including tulip_output and xsec.

The microscopic XS of each case will be saved in xsec and named MATx (x stands for number of case). The following core calculation will find the few-group cross sections in xsec directory.

In addition, there are 5 output files generated after running TULIP. The location of output files and corresponding contained information are listed as:

File name Information Location

run_log.txt run log file tulip_output/casename summary.out summary of partial microscopic XS and flux tulip_output/casename regionX-binary_xs microscopic XS for core calculation tulip_output/casename regionX-dntr macroscopic XS for core calculation tulip_output/casename check_log.txt input check current directory

Sample run_log.txt:


38

Sample summary.out:


39

8.2 Sample LAVENDER output

When finishing LAVENDER calculation, lavender_output directory will be created. There are 4 output files generated after running LAVENDER. The output files and the contained information are listed as:

File name Information Location output.main main output file, include flux and keff, lavender_output/ output.dephist nuclide density and keff at each burnup step lavender_output/ output.reactivity results of reactivity calculation lavender_output/ *.vtk flux, power for each node, lavender_output/visual

Sample output.main:


40


41

9. Reference

1. Xianan Du, Liangzhi Cao, Youqi Zheng, et al., 2018. A hybrid method to generate few-group cross sections for fast reactor analysis. Accepted in Journal of Nuclear Science and Technology.

2. Youqi Zheng, Liang Qiao, Zi'an Zhai, et al., 2018. SARAX: A new code for fast reactor analysis part II: Verification, validation and uncertainty quantification. Nuclear Engineering and Design, 331, Pages 41-53.

3. Xianan Du, Liangzhi Cao, Youqi Zheng, et al. Solution of OECD/NEA SFRs benchmark using SARAX code system: Transient analysis. PHYSOR2018, Cancun, Mexico, 2018.

4. Xianan Du, Liangzhi Cao, Youqi Zheng, 2017. Method of generating homogenized fast reactor assembly constants based on point-wise cross section. High Power Laser and Particle Beams, 29(1), Pages 5-10. (In Chinese)

5. Youqi Zheng, Yunlong Xiao, Hongchun Wu, 2017. Application of the virtual density theory in fast reactor analysis based on the neutron transport calculation, Nuclear Engineering and Technology, 320, 200-206.

6. Xianan Du, Liangzhi Cao, Youqi Zheng. The generation of few-group constants for fast reactor analysis. International Conference on Nuclear Engineering 2016, Charlotte, USA, 2016.

7. Mingtao He, Hongchun Wu, Youqi Zheng*, et al., 2015. Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method, Nuclear Engineering and Design, 295, 489–499.

8. Shengcheng Zhou, Hongchun Wu, Liangzhi Cao, et al., 2014. LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors. Nuclear Engineering and Design, 278, 434–444.

9. Xianan Du, Liangzhi Cao, Youqi Zheng, 2014. Application of Monte Carlo Method in Fast Reactor Assembly Homogeneous Constant Calculation. Nuclear Power Engineering, 35, Pages 67-70. (In Chinese)

10. Haoliang Lu, Hongchun Wu, 2007. A nodal SN transport method for three-dimensional triangular-z geometry, Annals of Nuclear Energy 237, 830-839.


42

Appendix A List of Isotope Symbol in TULIP code Considering the Doppler Effect of different isotope are distinguishing, the number of temperature point is not the same. There are 3 kinds when producing library.

1 point: Only 300 K cross sections are prepared in library files; 3 point: 300 K, 600 K, and 1000 K cross sections are prepared; 5 point: 300 K, 600 K, 1000 K, 1500 K, and 3000 K cross sections are prepared.

Isotope Temp. Point Isotope Temp. Point Isotope Temp. Point

Ag107 1 Ge76 1 Ru105 1

Ag109 1 H1 3 Ru106 3

Ag110m 1 H2 3 S32 1

Ag111 1 H3 3 S33 1

Al27 3 He3 1 S34 1

Am241 5 He4 1 S36 1

Am242 5 Hf174 1 Sb121 3

Am242m 5 Hf176 1 Sb123 1

Am243 5 Hf177 1 Sb124 1

Am244m 5 Hf178 1 Sb125 3

Am244 5 Hf179 1 Sb126 1

Ar36 1 Hf180 1 Se74 1

Ar38 1 I127 3 Se76 1

Ar40 1 I129 3 Se77 1

As74 1 I130 3 Se78 1

As75 1 I131 3 Se79 1

Au197 1 I135 1 Se80 1

B10 1 In113 1 Se82 1

B11 1 In115 1 Si28 1

Ba130 1 K39 1 Si29 1

Ba132 1 K40 1 Si30 1

Ba133 1 K41 1 Sm144 1

Ba134 1 Kr78 1 Sm147 1

Ba135 1 Kr80 1 Sm148 1

Ba136 1 Kr82 1 Sm149 1

Ba137 1 Kr83 3 Sm150 3

Ba138 3 Kr84 1 Sm151 1

Ba140 1 Kr85 1 Sm152 1

Be7 1 Kr86 1 Sm153 1

Be9 1 La138 3 Sm154 3

Bi209 1 La139 3 Sn112 1

Bk249 1 La140 3 Sn113 1

Bk250 1 Li6 1 Sn114 1

Br79 1 Li7 1 Sn115 1


43


Br81 1 Lu175 1 Sn116 1

C12 1 Lu176 1 Sn117 1

Ca40 3 Mg24 1 Sn118 1

Ca42 3 Mg25 1 Sn119 1

Ca43 3 Mg26 1 Sn120 1

Ca44 3 Mn55 3 Sn122 1

Ca46 3 Mo92 1 Sn123 1

Ca48 3 Mo94 1 Sn124 1

Cd106 1 Mo95 1 Sn125 1

Cd108 1 Mo96 1 Sn126 1

Cd110 1 Mo97 1 Ta181 1

Cd111 1 Mo98 1 Ta182 1

Cd112 1 Mo99 1 Tc99 3

Cd113 1 Mo100 1 Te120 1

Cd114 1 N14 3 Te122 1

Cd115m 1 N15 3 Te123 1

Cd116 1 Na22 1 Te124 1

Ce136 1 Na23 1 Te125 1

Ce138 1 Nb93 3 Te126 1

Ce139 1 Nb94 3 Te127m 3

Ce140 3 Nb95 3 Te128 3

Ce141 3 Nd142 1 Te129m 1

Ce142 3 Nd143 3 Te130 3

Ce143 1 Nd144 3 Te132 1

Ce144 3 Nd145 3 Th227 5

Cf249 1 Nd146 3 Th228 5

Cf250 1 Nd147 3 Th229 5

Cf251 1 Nd148 3 Th230 5

Cf252 1 Nd150 3 Th232 5

Cf253 1 Ni58 1 Th233 5

Cf254 1 Ni59 1 Th234 5

Cl35 1 Ni60 1 Ti46 1

Cl37 1 Ni61 1 Ti47 1

Cm241 5 Ni62 1 Ti48 1

Cm242 5 Ni64 1 Ti49 1

Cm243 5 Np235 5 Ti50 1

Cm244 5 Np236 5 U232 5

Cm245 5 Np237 5 U233 5

Cm246 5 Np238 5 U234 5

Cm247 5 Np239 5 U235 5

Cm248 5 O16 3 U236 5

Cm249 5 O17 3 U237 5

Cm250 5 P31 1 U238 5


44


Co59 1 Pa231 5 U239 5

Cr50 1 Pa232 5 U240 5

Cr52 1 Pa233 5 U241 5

Cr53 1 Pb204 1 Vnat 1

Cr54 1 Pb206 1 W182 3

Cs133 1 Pb207 1 W183 3

Cs134 3 Pb208 1 W184 3

Cs135 1 Pd102 3 W186 3

Cs136 1 Pd104 1 Xe123 1

Cs137 3 Pd105 3 Xe124 1

Cu63 1 Pd106 3 Xe126 1

Cu65 1 Pd107 1 Xe128 1

Dy156 1 Pd108 3 Xe129 1

Dy158 1 Pd110 3 Xe130 1

Dy160 1 Pm147 3 Xe131 1

Dy161 1 Pm148m 3 Xe132 3

Dy162 1 Pm148 3 Xe133 3

Dy163 1 Pm149 3 Xe134 3

Dy164 1 Pm151 1 Xe135 1

Eu151 1 Pr141 3 Xe136 3

Eu152 1 Pr142 3 Y89 1

Eu153 1 Pr143 3 Y90 1

Eu154 3 Pu236 5 Y91 3

Eu155 3 Pu237 5 Zr90 1

Eu156 1 Pu238 5 Zr91 1

Eu157 1 Pu239 5 Zr92 1

F19 3 Pu240 5 Zr93 1

Fe54 1 Pu241 5 Zr94 1

Fe56 3 Pu242 5 Zr95 1

Fe57 1 Pu243 5 Zr96 1

Fe58 1 Pu244 5 U235fp 1

Ga69 1 Pu246 5 U236fp 1

Ga71 1 Rb85 3 U238fp 1

Gd152 1 Rb86 1 Np237fp 1

Gd153 1 Rb87 3 Np239fp 1

Gd154 1 Rh103 1 Pu238fp 1

Gd155 1 Rh105 3 Pu239fp 1

Gd156 1 Ru96 3 Pu240fp 1

Gd157 1 Ru98 3 Pu241fp 1

Gd158 1 Ru99 3 Pu242fp 1

Gd160 1 Ru100 3 Am241fp 1

Ge70 1 Ru101 1

Ge72 1 Ru102 3


45


Ge73 1 Ru103 3

Ge74 1 Ru104 3


46

Appendix B Interface to SN code using ANISN format Only 1968-group cross sections can be stored in ANISN format. 1968×1972 cross sections are written in a file with decimal format. The arrangement of cross sections is as follow:

gχ ,a gΣ ,f gυΣ ,t gΣ

Example


47

Appendix C Interface to Use Hybrid Method

Example: Nuclide 1 Nuclide 2

Total Fission Elastic Total Fission Elastic

6.32452E+00 2.17948E+00 3.51161E+00 1.84459E+00 0.00000E+00 9.90067E-01 Group1

6.30758E+00 2.18251E+00 3.48915E+00 1.84083E+00 0.00000E+00 9.83296E-01 Group2

6.29844E+00 2.18433E+00 3.47675E+00 1.83721E+00 0.00000E+00 9.76777E-01 Group3

6.28347E+00 2.18760E+00 3.45618E+00 1.83336E+00 0.00000E+00 9.69851E-01 Group4

……

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