Abstract The use of advanced uranium-based and thorium-based fuel bundles in a 700-MW(electric)–class pressure tube heavy water reactor (PT-HWR) has the potential for improved performance characteristics with higher burnup, higher fissile fuel utilization, and lower coolant void reactivity while also extracting the energy potential in thorium. In this study, thermal-hydraulic subchannel analyses were

Abstract Throughout the years, various reports and training manuals on CANada Deuterium Uranium (CANDU) reactors have mentioned that the CANDU lattice is overmoderated. Overmoderation is not always defined in such documents but often appears associated with the positive void reactivity of the CANDU lattice. Some documents refer, logically, to overmoderation as meaning that the lattice pitch is larger

Abstract Improved methods for storing liquid hydrogen in larger quantities and over longer periods of time in space are becoming progressively more critical as sights are once again set on Mars. Current storage methods involve the venting of vaporized hydrogen to space, with the consequence that significant amounts of hydrogen are wasted. Extra hydrogen must be stored to account for this loss resulting

Abstract The purpose of this paper is to analyze how competence in the humanities and social sciences has been introduced into the system culture of the Swedish nuclear waste system (SNWS) traditionally dominated by scientists and engineers. In the spring of 1980, fierce local protests were directed against drilling teams sent out to investigate the geology of potential locations for a repository of

Abstract Ceramic-metallic nuclear fuels are a candidate fuel for nuclear thermal propulsion systems due to their high heat transport properties, which are necessary in very high-temperature environments. The conventional fabrication of uranium nitride–molybdenum fuel has been thoroughly studied in the past, but modern manufacturing techniques have presented a unique opportunity for further development

(2020). Technical Reviewers. Nuclear Technology: Vol. 206, No. 12, pp. iii-vii.

Abstract Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal-hydraulic analysis were coupled into an integrated code system. In the system, each code provides data required by other codes, and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical, and mechanical behaviors in a fuel

Abstract The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Earlier lattice physics and reactor core physics studies have demonstrated the feasibility of using such advanced fuels; however

Abstract The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Previous lattice physics and core physics studies have demonstrated the feasibility of using such advanced fuels; however, thermal-hydraulic

(2020). Foreword: Special issue on Salt-Cooled Reactors. Nuclear Technology: Vol. 206, Special issue on Salt-Cooled Reactors, pp. iii-iv.

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 proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed technique employs a multivariant analysis technique, namely, principal component analysis (PCA) and clustering analysis (CA) to identify the position of the PWR control rod using its impact on the core

Abstract This paper outlines research on understanding, characterizing, and managing conservatisms in safety analyses. This research includes a review of national and international approaches for developing and using conservative and best-estimate analyses. A probabilistic approach is discussed to support reducing conservatism while maintaining safety margins. An example of the proposed approach is

Abstract Seismic analysis and risk assessment of safety-critical infrastructure like hospitals, nuclear power plants, dams, and facilities handling radioactive materials involve computationally intensive numerical models and coupled multiphysics scenarios. They are also performed in a strict regulatory environment that requires high software quality assurance standards, and in the case of safety-related

Abstract In a companion paper, the U.S. Nuclear Regulatory Commission (NRC) staff has described analyses performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code to study the transient system response of the NuScale power module to a postulated beyond-design-basis loss of alternating-current (LOAC) power transient where the module protection system completely fails to insert the control

Abstract The U.S. Nuclear Regulatory Commission (NRC) staff often performs confirmatory analysis to support regulatory decision making. In the current work the TRAC/RELAP Advanced Computational Engine (TRACE) code was used to study the transient system response for the NuScale power module to a beyond-design-basis event where the control rods fail to insert. The regulatory purpose of the current work

Abstract One of the most important considerations in the design of electronic systems for post-accident monitoring in a nuclear power plant is how to deal with the complex and uncertain radiation environments. Without using special design methodologies and adequate protection, nonradiation-hardened commercial-off-the-shelf (COTS) electronic components can easily be damaged. In this paper, a new design

Abstract Accident tolerant fuels (ATFs) are new nuclear fuels developed in response to the accident at the Fukushima power station in March 2011. The goal of ATFs is to withstand accident scenarios through better performance compared to currently employed fuels (e.g., small-scale hydrogen generation). This paper targets a method for evaluating and comparing ATF performance from a probabilistic risk

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 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.

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