This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un'SCRAM'ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.

A nuclear reactor operates in an environment where complex multi-physics and multi-scale phenomena exist, and it requires consideration of coupling among neutronics, thermal hydraulics, fuel performance, chemical dynamics, and coupling between the reactor core and first circuit. Safe, reliable, and economical operation can be achieved by leveraging high-fidelity numerical simulation, and proper considerations for coupling among different physics and required to provide powerful numerical simulation tools. In the past simplistic models for some of the physics phenomena are used, with the recent development of advanced numerical methods, software design, and high-performance computing power, the appeal of multi-physics and multi-scale modeling and simulation has been broadened.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V & V) of available analytical capabilities for HTGR simulation for design and safety evaluations [1], [2], [3]. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.

This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and recriticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios.During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. Inthe multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level.In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling.

This book provides a systematic and comprehensive introduction to the neutronics of advanced nuclear systems, covering all key aspects, from the fundamental theories and methodologies to a wide range of advanced nuclear system designs and experiments. It is the first-ever book focusing on the neutronics of advanced nuclear systems in the world. Compared with traditional nuclear systems, advanced nuclear systems are characterized by more complex geometry and nuclear physics, and pose new challenges in terms of neutronics. Based on the achievements and experiences of the author and his team over the past few decades, the book focuses on the neutronics characteristics of advanced nuclear systems and introduces novel neutron transport methodologies for complex systems, high-fidelity calculation software for nuclear design and safety evaluation, and high-intensity neutron source and technologies for neutronics experiments. At the same time, it describes the development of various neutronics designs for advanced nuclear systems, including neutronics design for ITER, CLEAR and FDS series reactors. The book not only summarizes the progress and achievements of the author’s research work, but also highlights the latest advances and investigates the forefront of the field and the road ahead.

This is the third in a series of three volumes of proceedings of the 23rd Pacific Basin Nuclear Conference (PBNC 2022) which was held by Chinese Nuclear Society. As one in the most important and influential conference series of nuclear science and technology, the 23rd PBNC was held in Beijing and Chengdu, China in 2022 with the theme “Nuclear Innovation for Zero-carbon Future”. For taking solid steps toward the goals of achieving peak carbon emissions and carbon neutrality, future-oriented nuclear energy should be developed in an innovative way for meeting global energy demands and coordinating the deployment mechanism. It brought together outstanding nuclear scientists and technical experts, senior industry executives, senior government officials and international energy organization leaders from all across the world. The proceedings highlight the latest scientific, technological and industrial advances in Nuclear Safety and Security, Operations and Maintenance, New Builds, Waste Management, Spent Fuel, Decommissioning, Supply Capability and Quality Management, Fuel Cycles, Digital Reactor and New Technology, Innovative Reactors and New Applications, Irradiation Effects, Public Acceptance and Education, Economics, Medical and Biological Applications, and also the student program that intends to raise students’ awareness in fully engaging in this career and keep them updated on the current situation and future trends. These proceedings are not only a good summary of the new developments nuclear science and technology, but also a useful guideline for the researchers, engineers and graduate students.

The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle the larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional "operator split" approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.