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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Graphite for Advanced Nuclear Reactors: Deployment Readiness Review

Historically, graphite has been used in numerous reactor technologies, including research/test reactors and, commercially, in advanced gas reactors. These non-metallic materials can play a key role as internal core structures, reflectors, or neutron moderators, making them important for deployment of certain advanced reactor technologies. This report explores the industry readiness for graphite material deployment, documenting Codes and Standards applicable to their design, qualification, and manufacturing. Discussions also examine the manufacturing processes, aging/degradation, inspection techniques, and disposal options. The report covers the current status, documents gaps, and proposes some conclusions about approaches to managing some of the current gaps.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A Refined BISON-Griffin Workflow For Designing Targeted TRISO Transient Experiments In The Transient Reactor Test Facility

U.S. industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In our previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs,(2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. Predicted tangential stresses in particle silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. Use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

PSA 2025 Presentation: "Modeling and Sensitivity Analysis of a Generation IV Pebble Bed Reactor Using MELCOR 2.2"

Accompanying the advancement of reactor technologies is the need for computational modeling and simulation to predict their behavior under normal operating conditions and accident scenarios. New Generation IV reactor designs which employ non-conventional fuel have a particular need for modeling the behavior and release of radionuclides and other material from the fuel. In this work, MELCOR version 2.2, a system-level safety and accident scenario code developed by Sandia National Laboratories, was used to model a 200-MWth pebble bed modular reactor and calculate the inventories of circulating and deposited graphite, metal dust, and elemental components released from the fuel elements. A base case modeling the reactor under standard operating conditions was calculated using MELCOR and the inventories were extrapolated to 30 years of operation time using a logarithmic regression fit. A sensitivity analysis was also performed in which several key parameters for the base case model were modified to explore the effect of these changes on the inventories calculated by MELCOR. A set of transient scenario simulations for a depressurized loss of forced cooling (DLOFC) accident were also performed. The results of the sensitivity analysis and transient simulations are reported and discussed in relation to the modeling techniques used for this study.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Development of Advanced Remote Cutting Technology for Reactor Internals Segmentation - 18274

Since the Three Mile Island decommissioning in the 1980's, Westinghouse has been involved in dismantling and segmentation of commercial reactors. With dozens of components segmented and packaged across the globe, Westinghouse has continued to evaluate the state of the art cutting technologies for their applications in the nuclear industry. While there have been well documented issues of certain cutting technologies in field applications, there are advancements of these technologies which prove them applicable for specific areas of the decommissioning process. During 2017, Westinghouse embarked on a revision of its plasma arc cutting (PAC) technology as part of an extensive innovation program for reactor internals segmentation. While these systems were used in Yankee Rowe and Three Mile Island, new plasma arc cutting systems improve the versatility and have much better precision of cutting. While bringing this technology into the future, Westinghouse is adapting the mounting and application so that it is compatible with the existing mechanical cutting portfolio. This means that in the event that a mechanical cutting technique is unable to perform as intended, a plasma cutting alternative will be available with minimal time to switch cutting methods. A test program for underwater PAC was performed, and the test gave a good reference of cutting speed and handling in different plate thicknesses which can be used when evaluating the use of the new cutting technology for future projects. PAC technology has proven that it could have a future in reactor vessel internals segmentation projects as a complement to the existing mechanical cutting methods under water. Thicknesses up to at least 25 mm can be cut with a much higher speed than the existing mechanical cutting methods in use. This and other cutting technology developments as part of the innovation program provide exciting new possibilities for dismantling and segmentation projects. While each of the techniques presents definite benefits for operators and, more importantly, customers, there are risks that need to be mitigated through engineered processes and controls. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

The coupled neutronics and thermo-fluid dynamics design characteristics of the XE-100 200 MW{sub TH} reactor - 18552

The Xe-100 is a high temperature helium-cooled, graphite moderated, pebble bed reactor with a 10 times through, multi-pass (MEDUL=MEhrfacDUrchLauf) fueling scheme. The design, characterized as a Generation IV reactor, is based on proven technology that has been demonstrated a number of times both experimentally and as commercial reactors. This proven foundation ensures high levels of design and technical readiness. In the overview provided of the design it is shown how the excess reactivity is limited by continuous refueling, whilst the selected core power density will ensure passive heat removal. In this way the intrinsic safety design features of a typical GEN IV design are guaranteed. It is shown that no upset event will lead to a significant amount of radiological release. The coupled core neutronic and thermo-fluid dynamics design for the Xe-100 is performed with the VSOP-A system of codes. For the equilibrium core, VSOP results demonstrate that the spherical fuel powers (maximum 1.9 kW) and operational temperatures (<900 deg. C) fall well within the envelope of the design criteria. Adequate reactivity control and long-term, cold shutdown are provided by two separate, diversely-actuated systems, while the overall negative reactivity temperature coefficient is illustrated over the total operational range. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236 U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individual component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

SuPer-Homogenization (SPH) Corrected Cross Section Generation for High Temperature Reactor

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy of full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Delivery of Dynamic Thermal Energy Storage Models and Advanced Reactor Concept Models to the HYBRID Repository

This publication details newly created energy storage and reactor models developed within the HYBRID modeling repository as part of the Department of Energy Office of Nuclear Energy (DOE-NE) Integrated Energy Systems (IES) program, led by Idaho National Laboratory (INL). Model development to-date includes creation of dynamic systems-level models of a pebble bed high temperature gas reactor (HTGR), sodium fast reactor (SFR), compressed air energy storage (CAES), liquid air energy storage (LAES) and Modelica standard library based two-tank sensible heat storage (SHS) in the IES-based HYBRID repository. Models are developed using the latest publicly available data and incorporate the possibility of control strategy inclusion for use with the existing IES modeling, analysis, and optimization toolset. Simulations showcase the abilities of each technology to flexibly operate in ways consistent with IES operation expectations. When these models are available, they can be utilized within different integrated energy park concepts to understand optimal system operation, control, and dispatching. Moreover, given the generic nature of the models, industrial partner technologies can be quickly added to the repository using the existing models as a basis. Additional dynamic models for thermal energy storage concepts can be developed and added to the HYBRID repository as needed. Also detailed in this report are future development goals for the HYBRID repository including adding suites of steady-state models, economic costing information, and reduced-order models. By adding these models in addition to the physical transient models currently existing within HYBRID, HYBRID will be a fully integrable tool for FORCE users.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Optimized design of reactor cores for continuous operation of Belgian nuclear power plants

Two Belgian nuclear power plants, owned and operated by Electrabel, were initially planned to be phased out in 2015, reaching their maximum life time of operation by the initial design. The fuel management of those two units has been subsequently designed by Tractebel to make the best use of the energy in the fuel assemblies either in the core or in the storage pool and to minimize the size of the fresh reloads in the terminating cycles. However, for ensuring energy security in Belgium, the Government decided in the same year to extend the life time of operation of the two units for 10 years, which raised the need to resume their continuous operation, in the opposite direction of the core contents and with a significant break in the usual sequence of cycle reloads. All this required adaptation of the tools and process for optimized reactor core design, in order to ensure the continuous and flexible operation of those units. The Tractebel LWRWIMS-PANTHER (LWP) core physics package is based on the 3-dimensional (3D) neutronic code PANTHER and the lattice code LWRWIMS for the generation of the assembly nuclear data. Both have been qualified by Tractebel for the Belgian Units to perform a complete range of PWR reactor calculations for fuel management design safety parameters assessment, transient analysis and operational support. In-core fuel management involves calculating core reactivity, power distribution and isotopic inventory for subsequent fuel cycles of a nuclear power plant in order to maintain adequate safety margins and operating lifetime for each core loading. This paper presents Tractebel's improved core design tools and process, the approach to optimize the reactor core design, and the applications to the 2 successive but opposite design requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Radiological characterization and Decommissioning of Research and Power Reactors - 15602

The production of the detailed activity and dose rate atlases (ADAs) is one of the indispensable measures in the decommissioning process of nuclear reactors which allows to determine the radiation field for optimal radiation protection as well as to quantify and characterize nuclear waste for disposal. High-detailed simulation models capable of performing neutron and radiation transport calculations at different levels of precision and details are applied for this purpose. Within the framework of the research project 'Decommissioning of Research and Power Reactors', funded by the German Federal Ministry of Education and Research (BMBF), and implemented by the Institute of Nuclear Engineering and Technology Transfer (NET) -RWTH Aachen, sophisticated simulation tools are developed. The enhances capability of these tools for the generation of the activity and dose rate distribution are tested and demonstrated on the shutdown German research reactor FRJ-2. To achieve the project goal different computational steps were required: 1. Development of a calculational CAD-based model for Monte Carlo neutron and gamma transport simulation using converter tool McCad 2. Design of the neutron source which reproduce the neutron distribution leaving the core 3. Enhancement of the simulation performance by using the hybrid deterministic Monte Carlo variance reduction methods 4. Calculation of N-flux distribution in the entire geometry 5. Development of couple codes between MCNP5 and the activation code ORIGEN2.2 This paper describes the main significant project milestones. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Lessons Learned from the Remote Reactor Vessel Internals Segmentation at Barsebaeck 2 - 18231

Westinghouse is performing a decommissioning project at Barsebaeck in Sweden that was awarded in June 2015. Barsebaeck is an ABB-designed boiling water reactor (BWR) plant (2 units of 615 MW) located 30 kilometers from Malmoe. The first reactor, Barsebaeck 1, was permanently shut down on Nov. 30, 1999, and the second one, Barsebaeck 2, ceased operation on May 31, 2005. The scope of work covers the underwater segmentation and packaging of the Reactor Vessel internals at Barsebaeck 1 and 2, including the up-front engineering studies and equipment manufacturing and qualification. The Barsebaeck Unit 2 was selected for starting the site work in August 2016. Cutting was performed with different types of Westinghouse designed cutting equipment (hydraulic shears, band saw, disc saw). All segmentation tools are remotely controlled due to the high radiation levels and ALARA and personal safety is the number one priority during the site work. The cutting equipment proved to be reliable and flexible which is very valuable for the execution of such a challenging project. The Barsebaeck Unit 2 cutting activities have been finished end of October 2018 and its lessons learned will be taken into consideration for the Unit 1 activities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Near-Term Deployment Viability of Liquid-Fuel Molten Salt Reactors

Molten salt reactors (MSRs) with liquid fuel offer many benefits such as continuous fuel reprocessing, which provides unprecedented flexibility in fuel cycle selection, and excellent performance both in thermal efficiency and intrinsic safety. However, challenges such as the need for proliferation-resistant design and a high-temperature, corrosive environment for structural materials have hindered their development. New MSR concepts, such as the ThorCon Do-able Molten Salt Reactor, aim to simplify design and operation in order to cut costs and enable deployment in a relatively short period of time. The proposed methods for reducing the impact of longstanding issues on MSR deployment have incentivized further technical assessment of the reactor design. This study selectively examined the (1) thermal hydraulics, (2) neutronics, (3) primary safety systems, and (4) economics of a representative liquid-fuel MSR to better understand the technical and economic viability for near term deployment of this technology. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗