SERPENTVTT-Finland – Jaakko Leppänen
NJOYLos Alamos National Laboratory - Jeremy Conlin
-COMET code/methodology
- MRT Methodology & RAPID code system.
Georgia Tech – Farzad Rahnema
Virginia Tech – Alireza Haghighat
SARAX: A fast reactor neutronic codeXi’an Jiaotong University – Liangzhi Cao
NESTLE- New CapabilitiesUniversity of Tennessee - G. Ivan Maldonado and Ondrej Chvala
Purdue University - Hany S. Abdel-Khalik and Dongli Huang
SuperMC: Nuclear design and safety evaluationInstitute of Nuclear Energy Safety Technology,
Chinese Academy of Sciences - Bin Wu

Workshops Regulation:

Limit occupancy: no more than 30 and not less than 15 attendees per workshop and they will be accepted in a first come first serve basis, date for first registration will be announced.

Workshop Fee: 100 US$ recovery fee.

Schedule: Sunday April 22, 2018 and time will be announced.

Serpent - A Monte Carlo Reactor Physics Burnup Calculation Workshop
(4 hours)

Serpent is a multi-purpose three-dimensional continuous-energy Monte Carlo particle transport code, developed at VTT Technical Research Centre his workshop covers the recent development and future applications of the Serpent Monte Carlo code. Serpent has been developed at VTT Technical Research Centre of Finland since 2004, mainly for the purpose of reactor physics applications. The code has been used for generating group constants for deterministic transport codes, but also as a high-fidelity neutronics solver in coupled multi-physics simulations.

The development of a new Serpent-based computational framework "Kraken" was started at VTT in 2017. Kraken is essentially a modular platform that provides reduced-order and high-fidelity code sequences for coupled core physics calculations.

Topics to be covered

1. Introduction to Serpent.

2. Group constant generation for reduced-order calculations.

3. Coupled high-fidelity simulations for best-estimate analyses.

4. Case examples.


Serpent website --

NJOY Processing System Workshop
(4 hours)

Topics to be covered

1. Introduction to the ENDF file format.

2. Introduction to modules needed to create a continuous-energy neutron ACE file.

3. How to use the processed data in MCNP.

4. How to plot nuclear data with NJOY.


After enrollment, a week prior to PHYSOR, it will be sent to the participants an email informing them of things they should download before the workshop.

There will be access to wireless internet so that the participants can download various NJOY input decks and such.

Participants should bring their own computer and should have NJOY installed prior to the beginning of the class.

COMET Method /code Workshop
Farzad Rahnema, Georgia Tech,
(1.5 hours)

COMET (COarse-MEsh radiation Transport) is a hybrid continuous energy stochastic deterministic transport code based on the incident flux response expansion method. The COMET method has THE fidelity of stochastic methods and computational speed that is several orders of magnitude faster than both stochastic and deterministic transport methods. The code decomposes a large heterogeneous system such as a reactor core into a set of small local problems (e.g., fuel assemblies) and for each unique local problem in the system, a solution called response function is obtained by a stochastic solver (response function generator). The overall solution to the global (core) problem is then obtained by repeatedly generating local solutions via a linear superposition of responses for the unique local problems.

The neutronic version of COMET solves for the eigenvalue and the detailed fission density distribution (e.g., fuel pins) in the core with Cartesian (e.g., BWR, PWR, CANDU) as well as hexagonal geometries (e.g., HTTR, VHTR, HTR, ABTR). The coupled photon electron version is used for both detector response and radiotherapy calculations.

In this workshop, the methods for all three applications will be described and the code for the core calculations will be demonstrated.

MRT Methodology, RAPID code system Workshop
Alireza Haghighat -
(2.5 hours)

In the MRT methodology, the problem of interest is partitioned into stages based on its physics, and each stage is represented by a response function or set of coefficients. These stages are combined into a linear system of equations which are solved iteratively using the pre-calculated functions and/or coefficients.

RAPID (Real-time Analysis for Particle-transport In-situ Detection) is developed based on the MRT methodology that enables its real-time simulation capability. The current version of RAPID is capable of simulating nuclear systems such as spent fuel pools, spent fuel casks, and reactor cores. RAPID solves for pin-wise, axially-dependent fission density, critical/subcritical multiplication, and detector response.

Furthermore, recently a multi-user virtual reality system (VRS) has been developed that provides a web-based application for input preparation, real-time simulation, and output processing and visualization in a virtual environment. For an introduction, please view the following demo

Topics to be covered

1. MRT Methodology and Example Applications.

2. RAPID code system.

3. VRS-RAPID demonstration.

4. Hands-on use by participants.


There will be access to wireless internet so that the participants can have remote access to VRS-RAPID. The current version of VRS-RAPID is optimized for a Personal Computer using the Google Chrome browser, but it can be accessed through iPad, Tablet, etc. using any other browser.

To facilitate establishing individual accounts, participants are encouraged to contact Prof. Haghighat prior to the workshop.

SARAX – A Fast Reactor Neutronics Code Development and Validation
(4 hours)

SARAX is a code for fast reactor neutronics analysis in steady state and transient, developed at Xi’an Jiaotong University from 2013. Uncertainty from the nuclear cross sections can be quantified in the new version. Recent developments and validation works of the SARAX code will be discussed in this workshop. The non-commercial version will be released for free in this workshop, which contains executable files for steady state, depletion, point kinetics and keff uncertainty analysis.

Topics to be covered

1. Introduction to SARAX.

2. Methods for fast reactor neutronics analysis and uncertainty quantification.

3. Validation of SARAX for new fast reactor design.

4. Case examples.


The codes will be sent to participants in this workshop using USB disk. Participant should bring their own computer (Windows or Max OS, with no less than 4GB memory and 10 GB hard disk) if he/she would like to run the code during the workshop.

Pre-Release Overview of Latest Capabilities of the NESTLE Simulator and Application to Few-Group Cross-Section Uncertainty Quantification
(4 hours)

The NESTLE few-group 3D nodal core simulator was developed originally in the early 1990s at NC State University under the direction of Prof. Paul J. Turinsky and has been used widely over the last twenty years. A collaboration among the University of Tennessee, Oak Ridge National Laboratory, and NC State University during the past decade has led to a new and improved version of NESTLE written in modern Fortran and developed with modern software engineering practices and is planned to be released through RSICC in 2018. New features include a simplified input format, a drift-flux model for high slip thermal hydraulics, advanced depletion and isotope tracking, output files compatible with VISIT visualization software, Cartesian and hexagonal geometry, pin power reconstruction, n-group capabilities, and more. The new features have expanded NESTLE's usage from pressurized water reactors to new core models including boiling water reactors, small modular reactors, and fluoride salt cooled high temperature reactors. The purpose of this workshop is to introduce the new features of NESTLE, highlight its latest features, illustrate its integration with lattice physics codes, and demonstrate its ability to model and simulate a wide variety of contemporary and next generation reactor cores.

Furthermore, via collaboration with Purdue University, this workshop will also describe the application of the physics-guided coverage mapping (PCM) methodology, recently introduced to support model validation, to minimize the computational cost required for the preparation of the few-group cross-sections, while maintaining all the dependency details of the few-group cross-sections.

Topics to be covered

1. History, Background, Theory (45 minutes).

     a. Overview of new coding and software practices.

     b. Highlight of latest and most unique features.

     c. Lattice Physics Integration (Scale, Serpent, Casmo).

2. New Features and Representative Studies (45 minutes).

     a. PWR, SMR, BWR, VVER, FHR.

3. Efficient Evaluation of Core Simulator Few-Group Cross-Section Uncertainties via PCM (1 hour).

4. Hands-on Workshop (1.5 hour).

Registered attendees will be given the opportunity to edit and execute a few NESTLE models illustrative of the features and studies highlighted in this workshop.

SuperMC - Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation
(3 hours)

Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation, SuperMC, is a full-functional neutronics simulation software system, including radiation transport, depletion, activation and dose calculation. The advanced capabilities include CAD/image-based accurate modeling, intelligent data analysis based on multi-D/multi-style visualization and network collaborative nuclear analysis on cloud computing platform. It has been used in over 60 nations and more than 30 major nuclear engineering projects.

Topics to be covered

1. Introduction of advanced features and development process of SuperMC.

2. Introduction of the usage of SuperMC.

3. Practice of case teaching of scene experience.


The demo version of SuperMC will be distributed for free in this workshop. Participants can bring their own computers (Windows 7/10, with less than 4GB memory and 10 GB hard disk space) to install and run SuperMC during the workshop.

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