Super-COPD (JAEA)

This code is a one dimensional plant dynamics analysis code for Sodium Fast Reactors (SFRs), which was originally developed by JAEA for simulating whole plant and in-core thermal hydraulics. This code has been validated by many sodium test facilities in Oarai and the operation/test data of Joyo and Monju.

No open literature documentation is available for this code.

5.1.13. THACS (XJTU)

The 1-D systems analysis code THACS (Transient Thermal Hydraulic Code for Analysis of Sodium cooled fast reactor) has been developed by Xi’an Jiaotong University [36], [37]. The code can calculate single-phase and two-phase flows of sodium. One dimensional flow with non-compressibility is assumed in the single-phase flows of sodium. For the two-phase calculation, the multi-bubble model is used in the core module for sodium. A compressible water model is applied on the water side of steam generators. THACS uses an object-oriented structure permitting the representation of any kind of hydraulic circuit, from the analytical experimental facilities to the whole reactor power plant. The code capabilities are listed as below:

(a) Multiple channel core thermal hydraulic analysis;

(b) Point-kinetic resolution covering decay heat and reactivity feedback models, including fuel Doppler, fuel and coolant density variations, core radial expansion, control rod driveline expansion and coolant voiding;

(c) Models of metallic fuel and MOX fuel thermo-physical properties;

(d) Primary and intermediate loops of reactor coolant systems models, such as pipes, intermediate heat exchangers, centrifugal pumps, hot pools and cold pools, pipe-nets, air-dump heat exchangers, steam generators, inter-wrapper flow and reactor vessel auxiliary cooling systems.

The first version of THACS was published on 25, June 2014.

5.1.14. TRACE (KINS, NRG, PSI)

The TRACE (TRAC-RELAP Advanced Computational Engine) code [38] is the latest in a series of best estimate system codes developed by the US Nuclear Regulatory Commission (NRC) for analysing steady state and transient thermohydraulic-neutronic behaviour in light water reactors [39], as well as in advanced reactor systems cooled by helium, sodium or lead-bismuth eutectic [40]. It can also model phenomena occurring in experimental facilities designed to simulate transients in-reactor systems. Models used include multidimensional two-phase flow (single-phase flow for non-water fluids), non-equilibrium thermodynamics, generalized heat transfer, reflooding, level tracking and reactor kinetics, using either the point-kinetics model or the PARCS 3D [41] reactor kinetics solver integrated into TRACE.

5.2. NEUTRONICS CODES 5.2.1. DIF3D (Argonne)

DIF3D [42] solves one-, two-, and three-dimensional problems for neutron flux by applying either finite difference diffusion (FDD) theory or variational nodal Pn transport theory (VARIATN) solvers. The code is used to calculate the core keff and the power distribution, as well as to evaluate the core keff after radial expansion and axial expansion.

5.2.2. ERANOS (KIT, U. Fukui)

The ERANOS code [43] was developed at CEA (France) and validated within the European Collaboration on Fast Reactors in the 1980s with the aim of providing a suitable basis for reliable neutronics calculations of current and advanced nuclear reactors, with specific attention to fast neutron spectrum cores. The code is able to perform the overall neutronics analysis of 3-D, 2-D, and 1-D geometrical models from the multigroup neutron cross-section generation to the computation of direct and adjoint neutron flux distributions by solving the transport or diffusion equation. Several embedded functions allow performing burnup simulations, evaluation of the kinetics parameters, space-time kinetics and perturbation and sensitivity studies. The self-shielded neutron cross-sections are processed by means of the European Cell Code (ECCO) ([44], [45]) and several reference neutron data libraries, i.e.

JEFF or ENDF/B, may be employed.

5.2.3. MC²-3 (Argonne)

MC²-3 [46] is a multigroup cross-section generation code for fast reactor analysis. It is used to process neutron and photon cross-sections and to generate multigroup neutron and photon cross-sections. For photon power calculations, MC²-3 was also used to generate the matrix for transforming neutron flux to photon source, as well as the neutron and photon KERMA factors.

5.2.4. MCNP6 (ENEA)

MCNP6^TM is a general purpose, continuous energy, generalized-geometry, time-dependent Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies [47]. The code’s main capability is to calculate keff eigenvalues for fissile systems.

It is also able to perform material burnup and delayed particle production calculations.

Pointwise section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-VII) are accounted for. Thermal neutrons are described

by both the free gas and S(α,β) models. A flexible tally structure allows calculation of important parameters such as neutron flux distribution and fluences.

5.2.5. NJOY (U. Fukui)

NJOY [48] is a nuclear data processing system that generates neutron cross-sections from ENDF formatted evaluated nuclear data.

5.2.6. PARTISN (KIT)

The PARTISN (PARallel, TIme-dependent SN) code [49] is the successor to the DANTSYS code package [50]. PARTISN solves the time-dependent transport equation by using the SN method for 1-D, 2-D (RZ, XY, and R-θ), and 3-D (XYZ, R-Z-θ) geometries. In the past, the DANTSYS code was implemented as a transport solver in the SIMMER-III (2-D) and SIMMER-IV (3-D) multi-physics code systems (see Section 5.1.9). Since the original DANTSYS version solves steady state problems only, the code was extended at KIT [51] for simulating time-dependent problems while being implemented in SIMMER.

5.2.7. PHISICS (ENEA)

PHISICS [52] is an advanced neutronic simulation code. The internal neutron solver discretization scheme is based on the second-order PN equation with the hybrid finite element method also known as the Variational Nodal Method.The maximum angular flux expansion order is 33. Several node geometries such as Cartesian, hexagonal, unstructured triangular and so on are available. The number of energy groups is limited only by the hardware performance, and coupling with RELAP5-3D has just been implemented. The code is designed with the mind-set to maximize accuracy for a given availability of computational resources to provide a state of the art simulation capability to reactor designers.

5.2.8. SCALE6.1 (ENEA)

The Standardized Computer Analysis for Licensing Evaluation (SCALE) code system [53]

developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly toolset for criticality safety, reactor physics, spent fuel characterization, radiation shielding and sensitivity and uncertainty analysis. Since 1976, regulators, licensees and research institutions around the world have used SCALE for safety analysis and design. SCALE 6.1 provides improved reliability and introduces a number of enhanced features in a robust yet user-friendly package that are intended to improve safety and efficiency throughout the nuclear community.

5.2.9. SERPENT (Politecnico di Torino, PSI)

Serpent ([54], [55]) is a continuous energy Monte Carlo (MC) neutron transport code being developed for reactor physics applications, including criticality calculations, burnup and decay analyses and generation of few-group homogenized cross-sections for deterministic full core simulators. Serpent uses continuous energy ACE-formatted cross-section libraries.

Typically, Serpent outperforms general purpose MC codes due to the use of the Woodcock delta-tracking in a combination with a typical surface-to-surface ray-tracing in a neutron tracking routine [56] and the use of the unionized energy grid for all pointwise reaction cross sections [57].

5.2.10. TRAIN (KIT)

The TRAIN code is part of the C4P-TRAIN code and data system package developed at KIT [58]. C4P [59] is used at KIT/IKET to produce problem-oriented neutron cross-section libraries in the CCCC format (ISOTXS and BRKOXS), which are needed in particular for the SIMMER code. The problem-oriented libraries are produced from fine group “master”

libraries generated from evaluated nuclear data files by codes like NJOY [48]. TRAIN is a code that - in combination with neutron transport solvers - can be used for reactor physics calculations with C4P libraries and helps to benchmark these libraries.