Workshop Titles, Schedule and Location


  • September 28, 2014 (Sun.)
Time Title Room Organizer
8:00-12:00 5. New features and capabilities in the Serpent 2 Monte Carlo code Mizuho A(4F) J. Leppänen (VTT, Finland)
8:00-12:00 3. TEPCO’s Activity on the Investigation of Fukushima Daiichi Accident/Current status of Fukushima-Daiichi NPS and decommissioning process Mizuho B(4F) Tokyo Electric Power Company
8:00-12:00 6. Attila4MC - CAD integration, automated deterministic variance reduction, and GUI setup for MCNP Mizuho C(4F) Varian Medical Systems
8:00-12:00 2. Workshop on Accelerator-Driven System in PHYSOR 2014 Hiei(2F) T. Sugawara (JAEA, Japan), and C. H. Pyeon (KURRI, Japan)
13:30-17:35 1. Short Course on Uncertainty Characterization in Nuclear Calculations Mizuho A(4F) H. S. Abdel-Khalik (Purdue Univ., US)
13:30-17:35 4. Neutron noise techniques for reactor diagnostics Mizuho B(4F) I. Pazsit, C. Demaziere, and V. Dykin (Chalmers Univ., Sweden)
13:30-17:35 7. Hybrid particle transport methods for solving complex problems in real-time Mizuho C(4F) A. Haghighat (Virginia Tech (VT), US), F. Rahnema, and B. Petrovic (Georgia Tech (GT), US)
13:30-17:35 8. NESTLE 3D Nodal Core Simulator: An Overview of Latest Features and Capabilities Hiei(2F) G. I. Maldonado, N. P. Luciano, and S. D. Hart (Univ. of Tennessee, US)

1. Short Course on Uncertainty Characterization in Nuclear Calculations

Detail Program

Length: 3 hours


Prof. Hany S. Abdel-Khalik, Associate Professor, Purdue University


To realize the benefits of predictive science, a concerted effort of in the areas of uncertainty quantification, sensitivity analysis, and Data Assimilation, collectively referred to as uncertainty characterization, is essential. This workshop is presented in the form of a short course, intended for young researchers and practitioners with little background on these subjects, to help familiarize them with the importance and relevance of uncertainty characterization to nuclear engineering applications. We believe this is important as the role of the uncertainties is expected to be integral to the design, safety, and operation of existing as well as next generation reactors. In addition to covering the basics, an overview of the current state-of-the-art will be given with particular emphasis on the challenges pertaining to nuclear reactor modeling. Given the nature of uncertainty characterization algorithms, a wide variety of supporting mathematical analyses will be discussed including, vector spaces, projection methods, variational theory, Bayesian analysis, and Monte Carlo sampling methods.

2. Workshop on Accelerator-Driven System in PHYSOR 2014

Detail Program

  • TEF-P project (K. Tsujimoto, JAEA, Japan)
  • FREYA project: experiments and status (A. Kochetkov, SCK-CEN, Belgium)
  • YALINA experiment (Y. Gohar, ANL, US)
  • ADS design for disposing of USA spent nuclear fuel inventory (Y. Gohar, ANL, US)
  • MYRRHA project (G. Van den Eynde, SCK-CEN, Belgium)
  • China ADS project (C. Liu, CAS, China)
  • KUCA experiment (C. Pyeon, Kyoto Univ., Japan)
  • Free discussion


The workshop on the Accelerator-Driven System (ADS) in PHYSOR 2014 will be held on September 28, 2014. The main purpose of this workshop is to share the latest information on the ADS R&D from the viewpoint of reactor physics, especially, reactor physics experiments.

3. TEPCO’s Activity on the Investigation of Fukushima Daiichi Accident/Current status of Fukushima-Daiichi NPS and decommissioning process

Detail Program

  • TEPCO’s Activity on the Investigation of Fukushima Daiichi Accident
  • Current status of Fukushima-Daiichi NPS and decommissioning process


  • Shinya Mizokami (Tokyo Electric Power Company, Japan)
  • Masayuki Yamamoto (Tokyo Electric Power Company, Japan)


Progressions of severe accident in Fukushima Daiichi Unit 1, 2, and 3 are very complicated and many issues have been still unresolved. TEPCO lists up important unresolved issues during the severe accident and is trying to address these issues. In the first part, updated accident analysis and progression of understanding on severe accident is presented, including some resolved “unresolved issues”. For example, TEPCO found part of external feed-water by fire-engine was bypassed through seal of a pump in the system of feed water line, which has considerable impact on the degree of core damage.

In the second part, decommissioning status of Fukushima-Daiichi is presented. Currently, reactor (fuel debris) is cooled by external feed water but it leaks out from containment vessel and accumulates in the underground level of reactor, turbine, and auxiliary buildings. Countermeasures of this highly contaminated water become very important to minimize its volume. In order to start unloading of fuel debris, various developments including repair of leak tightness of the containment vessel, observation inside the containment vessel to locate fuel debris, and so on are important. These developments are undergoing as national projects in collaboration with other countries.

4. Neutron noise techniques for reactor diagnostics

Detail Program

1. Basic principles.

Origin of neutron fluctuations in zero power systems and power reactors. Domains of dominance, mathematical description, diagnostic value of the fluctuations.

2. Theory of power reactor noise.

The Langevin equation of power reactor noise. Linear theory of the noise induced by weak perturbations. Solution with the Green’s function technique: noise source, transfer function and induced neutron noise. Auto- and cross-power spectra of the neutron noise.

3. Noise source modeling.

Representation of the various perturbations: control rod vibration, local channel instability, boiling noise (propagating perturbations), temperature noise, core-barrel vibrations.

4. Calculation of the dynamic transfer function.

Noise equations in simple systems with analytical solutions. Calculation methods in realistic, inhomogeneous systems. The CoreSim program for numerical calculation of the transfer function or directly the neutron noise.

5. Principles of power reactor noise diagnostics.

Inversion of the noise equations: expressing the properties of the noise source from the properties of the measured neutron noise and the transfer properties of the system. Determining either system properties (BWR stability) or the characteristics of the perturbation (vibration amplitude, void fraction etc.).

6.Practical exercises:

  • Evaluation of neutron noise measurements taken in Swedish PWRs and BWRs. Calculation of neutron noise spectra. Demonstration of neutron diagnostics: characterizing core barrel vibrations in a PWR, determining steam velocity from in-core neutron detectors in BWRs.
  • Tutorial on the functioning and use of CoreSim in diagnostic problems.


  • Imre Pázsit
  • Christophe Demazière
  • Victor Dykin


The workshop deals with the various types of neutron fluctuations which arise in a nuclear reactor and explains how they can be used as a non-invasive method for diagnosing the system. It consists of a theory part, followed by practical exercises.

In the theory part the origin, mathematical description and the diagnostic value of the fluctuations is discussed. The workshop then concentrates on the neutron noise in power reactors, which carries information on both the reactor itself and on the technological processes, acting as “noise sources” (vibrations, boiling etc.), which give rise to the neutron noise. The induced noise is calculated as a convolution between the noise source and the transfer function of the system. It is shown how the noise sources can be represented in terms of cross section fluctuations, and how the transfer function can be calculated from the space-time dependent diffusion equations.

In the practical exercises, two options will be available either in parallel or sequentially. Option one: both simulated data as well as measurements taken in Swedish PWRs and BWRs will be analysed by first calculating the auto- and cross power spectral density of the noise and then showing how they can be used for diagnosing the “health status” of either the core or that of the noise sources (severity of vibrations, localization of vibrating control rods or channel instabilities, determining void velocity and void fraction etc). Option two: a tutorial on the use of the dynamic core simulator CoreSim will be given. It will be demonstrated how it can be used to calculate the neutron noise by various perturbations in a real heterogeneous reactor and how such calculations can help determining important safety parameters from noise measurements.

The workshop material will be distributed as an interactive Mathematica notebook, or its cdf version for those who do not have access to Mathematica. The CoreSim tool will be available to participants upon request.

5. New features and capabilities in the Serpent 2 Monte Carlo code


Jaakko Leppänen, Maria Pusa, Tuomas Viitanen, Ville Valtavirta and Toni Kaltiaisenaho from Serpent developer team, presentations from the Serpent user community


Serpent is a Monte Carlo transport calculation code developed at VTT Technical Research Centre of Finland since 2004. The code is used for various reactor physics applications in more than 100 universities and research organizations around the world. This workshop presents the recently implemented features and capabilities in the development version of the code, Serpent 2, including automated calculation sequence for group constant generation, internal and external thermal hydraulics and fuel performance code coupling for multi-physics applications, and a photon transport mode for gamma heating and transport applications. The presentations enlighten the theory and methodology behind the new features, but also include examples and hands-on tutorials useful for new and experienced Serpent users.

6. Attila4MC - CAD integration, automated deterministic variance reduction, and GUI setup for MCNP

Detail Program

A demonstration, followed by a hands-on tutorial, will be provided in Attila4MC, an easy-to-use Graphical User Interface for rapidly setting up MCNP6 calculations.


Gregory Failla, Senior Product Manager - Attila Product Line


Attila4MC enhances user productivity by eliminating the most time consuming bottlenecks in running MCNP: CAD integration, variance reduction, and ease-of-use. Through the intuitive Attila4MC graphical user interface (GUI), analysts can now import geometry directly from CAD and leverage MCNP’s new unstructured mesh functionality. Attila4MC’s state-of-the-art deterministic solver automatically generates optimized weight windows with a minimal amount of user interaction, bypassing the need for less efficient, but time intensive manual variance reduction. Through Attila4MC’s GUI, most MCNP calculations can be set up entirely without editing an input deck, simplifying analysis and verification. SpaceClaim is offered as integrated part of Attila4MC, providing users with the full power of direct CAD modeling at their fingertips. With SpaceClaim, analysts can create, import, manipulate, and repair complex geometry in ways previously only possible by full-time dedicated CAD experts.

7. Hybrid particle transport methods for solving complex problems in real-time

Detail Program

  • Introduction to issues associated with the Monte Carlo eigenvalue calculations (Prof. Petrovic, GT) (15 min)

  • COMET methodology (Prof. Rahnema, GT) (35 min)

  • Multi-stage response function methodologies (Prof. Haghighat, VT) (40 min)

  • Demonstration of tools (45 min)
    1. COMET (Coarse Mesh Transport)
    2. INSPCT-S (Inspection of Nuclear Spent-fuel Pool Computation Tool – Spreadsheet version)


The goal of this workshop to introduce the audience to advanced 3-D radiation transport methodologies and tools for fast and accurate simulation of reactor core and spent fuel pools. For the latter focus will be on criticality safety and safeguards issues. The workshop will be comprised of two parts: i) discussion of advanced methodologies/tools developed at Georgia Tech and Virginia Tech; ii) demonstration of these tools.

These advanced methodologies can be referred to as response-function particle transport in which the problem of interest is partitioned into different stages that each can be represented by a response function or set of coefficients. These stages are combined into a linear system of equations that are solved iteratively using the pre-calculated functions and/or coefficients. To determine these functions or coefficients, a set of fixed source forward Monte Carlo and forward/adjoint deterministic calculations are performed for different material compositions and physical/geometric conditions.

For demonstration, the COMET (Coarse Mesh Transport) will be used of core physics calculations and the INSPCT-S (Inspection of Nuclear Spent fuel-Pool Calculation Tool ver. Spreadsheet) tool will be used for simulation of a spent fuel pool in real time while preserving accuracy of 3-D transport calculations. These code systems and related tools/codes will be demonstrated during the workshop.

8. NESTLE 3D Nodal Core Simulator: An Overview of Latest Features and Capabilities

Detail Program

  1. History, Background, Theory
  1. Overview of new coding and software practices
  2. Highlight of latest and most unique features
  1. New Features and Representative Studies
  1. PWR, SMR
  2. BWR
  3. VVER
  4. FHR
  5. Multi-cycle Fuel Optimization
  1. Lattice Physics Integration
  1. Hands-on Workshop
Registered attendees will be given the opportunity to edit and execute a few NESTLE models illustrative of the features and studies highlighted under the “New Features and Representative Studies”


G. Ivan Maldonado and Nicholas P. Luciano (Univ. of Tennessee, Knoxville, USA)


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 last five years has led to a new and improved version of NESTLE written in modern Fortran and developed with modern software engineering practices. New features include a simplified input format, a drift-flux model for high slip thermal hydraulics, advanced depletion and isotope tracking using ORIGEN, output files compatible with VISIT visualization software, and more. The new features have expanded NESTLE’s usage from large 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 NESTLE, highlight its latest features, illustrate its integration with lattice physics calculations, and demonstrate its ability to model and simulate a wide variety of contemporary and next generation reactor cores. NESTLE runs on Linux, Windows, and Mac OSX platforms.