Abstract Oak Ridge National Laboratory (ORNL) successfully demonstrated the Versatile Remediation Module (VRM), a prototype module designed and built by ORNL for on-site remote repair of welded stainless steel storage containers for spent nuclear fuel and high-level radioactive waste. This paper describes the VRM prototype and its design features and components to support continued long-term storage

Abstract To investigate thermal-hydraulic phenomena during an intermediate-break loss-of-coolant accident (IBLOCA) and evaluate the effect of a direct vessel injection (DVI) line break, an integral effect test using the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test facility was conducted as the B3.2 test item of the international cooperation project Organisation for Economic

Abstract Measurements and data are reported for the total hemispherical emissivity of Grade 91 steel [ASTM International (ASTM) A387 Grade 91] for the temperature range of 400 K to 1048 K using ASTM standard C835-06. The surfaces studied included (1) an electric discharge machining (EDM) cut, (2) Grade 91 steel sandblasted with 320-grit-sized alumina beads, (3) EDM-cut Grade 91 steel oxidized in air

Abstract Design evaluations and thermal-hydraulic predictions were undertaken to compare three candidate options for loss-of-coolant accident (LOCA)–testing capabilities at the Transient Reactor Test facility (TREAT). These options included a small water capsule, a large natural circulation capsule, and a forced-convection water loop. All options were found to create reasonable approximations of radial

Abstract Thermal stratification (TS) is a thermal-fluid phenomenon that can introduce large uncertainties to nuclear reactor safety. The stratified layers caused by TS can lead to temperature oscillations in the reactor core. They can also result in damages to both the reactor vessel and in-vessel components due to the growth of thermal fatigue cracks. More importantly, TS can impede the establishment

Abstract Chalk River Unidentified Deposit (CRUD) specimens were sampled and analyzed using an electron probe micro analyzer (EPMA; JEOL JXA-8230R) with a bundle of spent nuclear fuel (actual burnup 49 655 MWd/tonne U) from a Korean nuclear power plant. CRUD collection in the reactor refueling cavity was carried out using the following two methods. The first method used an Al2O3 scraper to scrape a

Abstract Austenitic alloys such as austenitic stainless steels and Ni-based alloys have been specified as container materials for molten salt reactors (MSRs). In MSR environments, these alloys are susceptible to carburization because (1) graphite components provide a source of carbon, (2) oxide films are not protective in molten halide salts, and (3) MSR operating temperatures fall within the temperature

Abstract This paper provides a comprehensive overview of the development of a NuScale power module in the Modelica process model ecosystem at Idaho National Laboratory (INL) as part of the U.S. Department of Energy’s Office of Nuclear Energy Integrated Energy Systems (IES) program. Model development has led to the creation of a dynamic NuScale module in the Modelica language that operates under natural

Abstract The structural integrity of a reactor pressure vessel (RPV) is a crucial issue for an operating nuclear power plant, especially in the beltline region, which suffers the highest neutron irradiation. Owing to its capability of considering parameters based on statistical distributions and provision of objective risk-informed results, the probabilistic fracture mechanics (PFM) method is widely

Abstract This paper discusses a new method of simulating the fuel assembly duct-bowing reactivity coefficient for EBR-II run 138B. Quantification of the fuel assembly duct-bowing reactivity effect in liquid metal–cooled fast reactors has been a persistent problem since they were first designed and operated. Simulation of the duct-bowing reactivity effect is difficult because the level of detail required

Abstract The detection and measurement of transuranic activity in ambient air by counting alpha particles is confounded due to the presence of short-lived alpha-emitting isotopes due to radon and thoron. This paper describes an algorithm intended for use in a gross alpha-based continuous air monitor. The algorithm is capable of generating the variation of gross alpha count rate with time from air filter

Abstract Following U.S. Department of Energy Order 420.1 C for the mitigation of natural phenomena hazards, such as earthquakes, to nuclear facilities through periodic reassessments, Idaho National Laboratory (INL) has developed the Seismic Hazard Periodic Re-Evaluation Methodology (SHPRM). The SHPRM involves seven criteria that evaluate changes to the seismic hazard at a site due to changes in the

Abstract Electricity markets are changing because of (1) the addition of wind and solar generating capacity and (2) the goal of a low-carbon electricity grid. The large-scale addition of wind and solar photovoltaics results in low wholesale electricity prices at times of high wind and solar output and high prices at times of low wind and solar input. Today, gas turbine combined cycle (GTCC) plants

Abstract The safety of the main control room in a nuclear power plant is an important research topic with practical implications. In this study, we used virtual reality technology and multimodal data to investigate the effect of experience on operators’ responses under emergency conditions. We asked participants to perform a series of tasks in a virtual fire emergency environment while simultaneously

Abstract The U.S. Nuclear Regulatory Commission (NRC) staff often perform confirmatory analyses using the TRAC/RELAP Advanced Computational Engine (TRACE) and Purdue Advanced Reactor Core Simulator (PARCS) codes to assist in regulatory decision making. Recently, the NRC staff have performed numerous such analyses of anticipated transient without SCRAM (ATWS) with core instability (ATWS-I) scenarios

(2020). Foreword: Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) Nuclear Technology: Vol. 206, Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18), pp. iii-iii.

Abstract Subchannel code analysis is one of the key thermal-hydraulic approaches for nuclear reactor design and safety analysis. At present, subchannel codes are employed to compute local thermal-hydraulic conditions on the rod bundle fuel assemblies of nuclear reactor cores and to predict the performance of nuclear cores during normal and hypothetical accident conditions. Currently, the subchannel

Abstract Even for small modular reactors (SMRs) with all large pipes removed, a small-break loss-of-coolant accident (SBLOCA) is an important design-basis accident (DBA). Experimental simulation of the SBLOCA scenario is essential before a prototype reactor is realized. The system-integrated modular advanced reactor (SMART) is one of the SMRs developed by the Korea Atomic Energy Research Institute

Abstract Transmission of information using elastic ultrasonic waves on existing metallic pipes provides an alternative communication option for a nuclear facility. The advantages of this approach consist of transmitting information through barriers, such as the containment building wall, with minimal modification of the existing hardware. Because bit rates on the order of kilobits per second are achievable

Abstract This study proposes an interpolation-based response surface surrogate methodology to manage a large number of scenarios in dynamic probabilistic risk assessment. It adopts the shape Dynamic Time Warping algorithm to cluster the interpolation neighborhood from time series sample data. The interpolation method was adapted from Taylor Kriging to allow a reduced-order model of the Taylor series

Abstract A new generation of dynamic methods has started receiving attention for nuclear reactor probabilistic risk assessment (PRA). These methods, which are commonly referred to as dynamic PRA (DPRA) methodologies, directly employ system simulators to evaluate the impact of timing and sequencing of events (e.g., failure of components) on accident progression. Compared to classical PRA (CPRA) methods

Abstract This work explores recent developments in severe accident analysis and risk assessment to inform and expand on these perspectives. Variations in nuclear reactor safety policy, reactor designs, extent of use of risk information in decision making, and other aspects can impact how safety policies regarding nuclear installations are developed and implemented. In particular, the relationship of

This study sets out to demonstrate the capability of using electrochemistry and laser-induced breakdown spectroscopy (LIBS) for concentration prediction of multiple species in a molten salt system at 773 K. Samples contained UCl3 ranging from 0 to 10 wt%, GdCl3 ranging from 0 to 5 wt%, and MgCl2 ranging from 0 to 1.5 wt%, with LiCl-KCl eutectic salt as the remainder. Multivariate models were produced

While general guidance for addressing individual elements of the key principles of risk-informed decision making (RIDM) for large commercial nuclear power plants is available in the literature, the implementation of RIDM can still be challenging, whether a mature RIDM framework exists or not. Traditionally, RIDM approaches have focused strongly on the use of risk information, particularly quantitative

The safety goals adopted by the U.S. Nuclear Regulatory Commission (NRC) consist of two qualitative safety goals backed up by two quantitative health objectives (QHOs). The QHOs establish risk limits for severe accidents in terms of their radiological consequences to affected individuals, in particular, the average individual health risks of early fatality and latent cancers from radiation exposure

This paper presents the approach of extending a classical generic event tree (ET) of a Level 2 Probabilistic Safety Analysis to the results of a probabilistic dynamic safety analysis. The example of creep-induced steam generator tube rupture has been chosen. The results of an Analysis of Thermal Hydraulics of Leaks and Transients with Core Degradation (ATHLET-CD)/Monte Carlo Dynamic Event Tree (MCDET)

(2020). Foreword: Selected papers from the 2019 Nuclear and Emerging Technologies for Space Topical Meeting (NETS 2019) Nuclear Technology: Vol. 206, Selected papers from the 2019 Nuclear and Emerging Technologies for Space Topical Meeting (NETS 2019), pp. iii-iii.

Nuclear thermal propulsion (NTP) could be an advanced technology to facilitate a new and excellent rocket engine that would at least double the performance of the best conventional chemical rocket engines. NTP has been under development for several decades and was selected as the leading candidate technique for the manned mission to Mars, as suggested in Design Reference Architecture 5.0. During development

The specific mass (or mass per unit power) is a fundamental performance metric in space power systems. For surface power, a low specific mass reduces launch costs and lander size. For nuclear electric propulsion, a low specific mass enables fast transit within the solar system. Studies on specific mass have typically focused on point designs and have not adequately explored the design space and scaling

This paper considers the issues of safety assessment of new nuclear power plant (NPP) projects with VVER Generation III+ reactors in relation to the probability target for large release, which is subject to verification in the development of a full-scale Probabilistic Safety Assessment (PSA) Level 2. The design solutions implemented in Generation III+ reactors allow reducing the probability of a severe

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment

Extension of the operational lifetime beyond 40 years is currently a noteworthy project in the field of nuclear safety in France, especially for 900-MW(electric) reactors that will be the first ones to go through the fourth periodic safety review. For 1300-MW(electric) reactors, the safety studies have already been engaged. Probabilistic Safety Assessments (PSAs) play an important role in this process

(2020). Foreword: Selected papers from the 2019 American Nuclear Society Student Conference. Nuclear Technology: Vol. 206, No. 7, pp. iii-iii.

The General Electric (GE®) PETtrace 860 cyclotron at the Missouri University Research Reactor (MURR) facility is used extensively for medical and research radioisotope production. However, no model exists of its performance, and the proton beam’s energy and spatial distribution are unmeasured. Here, an MCNP6 model was developed to improve upon the performance of the cyclotron target systems that are

(2020). Statement of Retraction: Visualization and Integration of Images of Radioactive Substances as Point Cloud Data in 3-D Environment Models. Nuclear Technology: Vol. 206, No. 7, pp. 1095-1095.

The electrochemical behavior of La2O3 was investigated in LiF-NaF-KF (FLiNaK, 46.5-11.5-42.0 mol %) eutectic at 700°C. In the electrochemical tests, two kinds of working electrodes, i.e., tungsten and graphite, were utilized. The present study showed that La3+ ions can be deposited in the form of La metal on a tungsten cathode or LaC2 on a graphite cathode, and O2− can be removed in the form of CO/CO2

In this paper, a set of stylized numerical benchmark problems is developed. These problems are based on the Oak Ridge National Laboratory preconceptual design of a fluoride-salt-cooled small modular advanced high-temperature reactor, or SmAHTR, that uses prismatic fuel assemblies with cylindrical pins/rods containing tri-isotropic fuel particles. A detailed description of the benchmark problems is

A new type of low-enriched uranium (LEU) fuel based on an alloy of uranium and molybdenum is expected to allow the conversion of U.S. domestic high-performance research and test reactors requiring high density fuel from highly enriched uranium (HEU) to LEU. The University of Missouri Research Reactor (MURR®) has undergone design and performance calculations for conversion to this LEU fuel. Presented

This study investigates the degradation of the heat transfer performance of a closed-circuit intermediate natural circulation heat transport loop used as a passive safety system in a nuclear power plant (NPP). The degradation arises from the strong thermal-hydraulic (TH) coupling of the loop operating characteristics, saturation temperature and pressure, and natural circulation flow rate, which determine

The pressurizer of a pressurized water reactor (PWR), as a spray-heating degasser, has been widely used to remove dissolved gas in the primary coolant of PWRs. In the real degassing process, the boundary conditions of the pressurizer may change, causing fluctuations in the degassing state and affecting the efficiency of degassing. However, open-published studies have focused mainly on the steady-state

Abstract In the event of a severe nuclear accident, one major concern is the release of radioactive material into the environment causing potential exposure of the general public to radiation. Among the volatile radionuclides are a range of tellurium isotopes. Due to the radioactivity and the volatility of tellurium, it has to be taken into account when assessing the overall effects of an accident

The standard approach in nuclear criticality safety analysis is to rely quite heavily—and in some cases exclusively—on passive controls, such as assuming all worst-case conditions are by default attained. This means assumptions are made such as no poison, optimum moderation, and pure fissile actinide content at the maximum mass with optimum full reflection. What is clearly attainable is something less

To study the burnup features of accelerator-driven subcritical systems (ADSs), simplified transmutation trajectories are imperative to make the simulation process more effective with acceptable precision. This process has long been considered a challenging task since the construction of simplified burnup chains often need complex judgments and experiences. Additionally, the burnup analysis of ADSs

The temperature field in a nuclear waste repository is an important issue with regard to the design and safety assessment of the repository. In this paper, a double-layer model for simulating the heat conduction near a single waste canister is established, and then, by applying the Laplace transform to the governing equations of the heat conduction in the buffer layer and the surrounding rock, the

Abstract The present U.S. nuclear fuel cycle faces challenges that hinder the expansion of nuclear energy technology. The U.S. Department of Energy identified four nuclear fuel cycle options that make nuclear energy technology more desirable. Successfully analyzing the transitions from the current fuel cycle to these promising fuel cycles requires a nuclear fuel cycle simulator that can predictively

Molten salts have been proposed as heat transfer media due to their superior thermal performance at elevated temperatures. A number of heat transfer correlations have been proposed in the literature for molten salts without explicitly considering the radiative heat transfer effect in the salts, which may not be negligible. This study therefore attempts to (1) quantitatively analyze the convective and

For several decades, extensive research has been performed on the recovery and purification of zirconium from spent nuclear fuel cladding using a variety of chlorination reaction processes. After the reaction between fuel cladding and chlorine gas, zirconium tetrachloride is separated from other chloride species based on their boiling/sublimation points; however, the presence of iron and niobium chloride

In order to develop a flow sheet for the purification procedure of a LiCl-KCl eutectic mixture with the underlying aim of avoiding Cl2 handling in an engineering-scale laboratory, the present work focuses on electrochemical investigations on vacuum dried eutectic mixtures LiCl-KCl + H2O and vacuum dried LiCl-KCl + H2O mixtures using cyclic voltammetry at 500°C. Complicated voltammetric features were

A whole-core, steady-state, thermal-hydraulic model for the cylindrical pin-type fluoride-salt-cooled small modular advanced high-temperature reactor (SmAHTR) is developed. In this preconceptual reactor design initially proposed by Oak Ridge National Laboratory, each fuel assembly in the graphite-moderated core has the FLiBe coolant flowing parallel to a hexagonal array of fuel and moderator pins.

A methodology for designing a Bonner sphere spectrometer was proposed based on the singular value decomposition (SVD) and information entropy theory. First, features of Bonner spheres were classified based on SVD, and then, the rationality of the selected spheres was analyzed. Second, the information entropy is used as the quantitative evaluation standard; combined with the unfolding method, the contribution

The Argonaut reactor is a water-moderated research reactor that was designed by the Argonne National Laboratory. There are currently few reactors of this type under operating conditions in the world, one of which, located at the Nuclear Engineering Institute in Brazil, has been in operation since 1965. This study analyzes a nonpostulated accident caused by the collision of the reactor coverer into

A proton therapy facility based on an isochronous superconducting cyclotron is under constructed at the Proton Therapy Facility of the Huazhong University of Science and Technology. Due to the limitation of minimum energy extracted by the energy selection system in the upstream beamline, a range shifter is installed at the end of the nozzle to further decrease the proton beam energy so that the shallow-seated

Analytical solutions for temperatures in an infinite region bounded internally by a cylinder have proved to be useful for thermal analysis of heat-producing nuclear waste disposal scenarios where the thermal design criteria are peak temperatures. The practicality of an analytical solution for the temperature of the host rock used in forced-ventilation thermal analyses has been illustrated by a computational

The optical properties of FLiBe salt in a Fluoride-Salt-Cooled High-Temperature Reactor (FHR) present an opportunity to utilize Cerenkov radiation measurements to reconstruct the power profile in the core and detect various anomalies that could occur during operation. The Cerenkov light produced within a coolant channel is strongly correlated to the fission rate density and power level in the surrounding

In order to model the activated isotopes and resulting dose from a nuclear detonation in an urban environment, the Activation and Transmutation of Isotopes in an Unstructured Mesh (ACTIUM) Python toolkit has been developed to combine the unstructured mesh–based particle transport capability of MCNP6.2 with the CINDER2008 transmutation code to produce quantities of interest for the post-detonation nuclear

Airborne contaminants from fires containing nuclear waste represent significant health hazards and shape the design and operation of nuclear facilities. Much of the data used to formulate DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities,” from the U.S. Department of Energy, were taken over 40 years ago. The objectives of this study were

A manual approach to radiography process optimization is a time-consuming and labor-intensive process. Therefore, a virtual environment in which all of the processes of optimization for a desired radiography experiment or setup are conducted is highly desirable. Such an environment should be able to provide the capability to arrive at radiographic scanning parameters that are optimized to within preset

Recent research has focused on the development of surrogate models for radiation source localization in a simulated urban domain. We employ the Monte Carlo N-Particle (MCNP) code to provide high-fidelity simulations of radiation transport within an urban domain. The model is constructed to employ a source location ( x , y , z ) as input and return the estimated count rate for a set of specified detector

Abstract During a severe accident in a nuclear reactor, there are a number of phenomenological events that can present a challenge to containment integrity. These include the generation and combustion of hydrogen, energetic fuel-coolant interactions, thermal attack of fission product barriers, core-concrete interactions, direct containment heating, and gradual overpressurization. The advanced design

Abstract In the framework of the European Union MAXSIMA project, the safety of the steam generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactor has been studied investigating the consequences and damage propagation of a SG tube rupture event and characterizing leak rates from typical cracks. Instrumentation able to promptly detect the presence of a crack in the SG tubes may