Advanced Reactor Designs Research Articles

Numerical study on the Welander oscillatory natural circulation problem using high-order numerical methods

In this paper, high-order numerical methods are investigated in a system analysis-like code. The classical Welander oscillatory natural circulation problem, which resembles a simplified example for many types of natural circulation loops widely seen in nuclear reactor systems, was chosen to illustrate the applicability of such methods in system analysis codes, and to demonstrate the advantages of such methods over the low-order methods widely used in existing system analysis codes. As originally studied by Welander, the fluid motion in a differentially heated fluid loop can exhibit stable, weakly unstable, and strongly unstable modes. A theoretical stability map has also been originally derived from the stability analysis. Numerical results obtained in this paper show very good agreement with Welander's theoretical derivations. For stable cases, numerical results from both the high-order and low-order numerical methods agree well with the non-dimensional flow rate that were analytically derived. The high-order numerical methods give much less numerical errors compared to those using low-order numerical methods. For stability analysis, the high-order numerical methods perfectly predicted the stability map even with coarse mesh and large time step, while the low-order numerical methods failed to do so unless very fine mesh and time step are used. The result obtained in this paper is a strong evidence for the benefits of using high-order numerical methods over the low-order ones, when they are applied to simulate natural circulation phenomenon that has already gained increasing interests in many existing and advanced nuclear reactor designs.

Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

New Safety Features in the Design of Advanced Nuclear Power Reactors

New Safety Features in the Design of Advanced Nuclear Power Reactors

Design of an Organic Simplified Nuclear Reactor

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Experimental observations of natural circulation flow in the NSTF

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary of some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states.

Recent development and application of a new safety analysis code for fusion reactors

This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this area and the application of this new capability to a DEMO relevant water-cooled tungsten armored divertor.

Subcritical space nuclear system without most movable control systems

This paper describes the design and analysis of advanced space nuclear reactor (ASNR) whose design combines the advantages of radioisotope thermoelectric generator (RTG) and space nuclear reactor (SNR). As opposed to current SNRs designs, ASNR is a subcritical system driven by 232U–Be neutron source to generate thermal power continuously. Most movable control systems in the SNR design are removed. The detailed neutronic calculations by MCNPX (Monte Carlo N-Particle eXtended), including keff, flux, burn-up, loss-ratio of neutron source and immersion reactivity, show that ASNR has higher criticality safety and more compact structure to bear the risk of immersion accident compared with the past SNRs, and the new system can provide more thermal power than RTG. Furthermore, the neutron source efficiency is optimized to improve the utilization of 232U–Be neutron source with the improvement of criticality safety. Compared with the past designs of space nuclear power, ASNR could provide enough thermal power and avoid the occurrence of serious immersion accident in the case of total control system failure. ASNR has potential for future deep space missions.

Progress in Nuclear Power Technologies and Implications for ASEAN

Since its inception, the nuclear industry has committed strongly to enhancing the safety features of Light Water Reactors (LWRs). Unfortunately, the series of nuclear accidents from Three Mile Island, to Chernobyl, and Fukushima led to spiking fear and strong negative perceptions of nuclear power. Since then, the nuclear industry has initiated a new wave of scientific and technological innovation towards safer LWR as well as advanced reactor designs. Besides the research and development of Generation III+ and IV types and thorium based reactors, recent years have also seen revived interests in fusion reactors. Given the early stage of development, the economic and environmental performance of the advanced reactor systems are yet to be evaluated. For the fast developing economies in the Association of South East Asian Nations (ASEAN), while environmental friendly energy sources are preferable, affordability of energy is critical. Given the resource and land constraints, several ASEAN members have expressed interest in nuclear power with Vietnam being the first one to build nuclear power reactors. However, the economic and environmental performance of the advanced reactors have not yet been determined due to their early stage of development. In addition, the safety and security of these advanced reactors combined with the under-developed governance on nuclear power are yet to be ascertained. Although some of the ASEAN members already have nuclear research programs, there is a lack of knowledge and expertise in civilian nuclear technology. If the advanced reactor technologies became commercially attractive, it is important to understand the implications for ASEAN's nuclear movement. Through analyzing the economic (commercial costs) and environmental (life cycle carbon emissions) performance of possible advanced reactors, I provide a preliminary competitive landscape for the advanced reactor systems. Based on the findings, I outline the policy considerations for ASEAN when advanced nuclear power reactors become commercially available. Essentially, this paper seeks to address an important question: what are the key steps in preparing the region towards “safer nuclear” in the long term.

CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria. Next, a pseudo transient simulation was carried out by taking the power and mass flow rate at different time instants to the Calandria from the 1D analysis, as an input to the 3D steady state simulations. The results indicate that temperature hotspots are located at the top center of the fuel channel matrix depending on the operating regime. Furthermore, it was observed that for a period of 7 days the temperatures are well within the safe limits.

Modeling defect cluster evolution in irradiated structural materials: Focus on comparing to high-resolution experimental characterization studies

Abstract

ICONE23-2083 FULL SCALE BWR CONTAINMENT LOCA RESPONSE TEST AT THE INKA TEST FACILITY

KERENA is an innovative boiling water reactor concept with passive safety systems (Generation III+) of AREVA. The reactor is an evolutionary design of operating BWRs (Generation II). In order to verify the functionality and performance of the KERENA safety concept required for the transient and accident management, the test facility "Integral Teststand Karlstein" (INKA) was built at Karlstein (Germany). It is a mock-up of the KERENA boiling water reactor containment, with integrated pressure suppression system. The complete chain of passive safety components is available. The passive components and the levels are represented in full scale. The volume scaling of the containment compartments is approximately 1:24. The reactor pressure vessel (RPV) is simulated via the steam accumulator of the Karlstein Large Valve Test Facility. This vessel provides an energy storage capacity of approximately 1/6 of the KERENA RPV and is supplied by a Benson boiler with a thermal power of 22 MW. With respect to the available power supply, the containment- and system-sizing of the facility is by far the largest one of its kind worldwide. From 2009 to 2012, several single component tests were conducted (Emergency Condenser, Containment Cooling Condenser, Core Flooding System etc.). On March 21st, 2013, the worldwide first large-scale only passively managed integral accident test of a boiling water reactor was simulated at INKA. The integral test measured the combined response of the KERENA passive safety systems to the postulated initiating event was the "Main Steam Line Break" (MSLB) inside the Containment with decay heat simulation. The results of the performed integral test (MSLB) showed that the passive safety systems alone are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them as response to an anticipated accident scenario. The test provided evidence that the INKA is worldwide the first large scale test facility to perform integral verification tests of passive safety concepts under plant-like scaling and thermodynamic conditions. Hence, the test facility also shows that it is capable to perform containment response tests for existing Generation II BWRs (with active safety systems) and advanced (passive) reactor designs besides KERENA. These test results can be used to strengthen existing containment codes with regard to heat transfer, natural circulation, gas- and temperature stratification and others.

A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing Fast Reactors

We developed a simulation tool that accelerates the evaluation of design changes on the equilibrium cycle of fast-spectrum nuclear reactors. Within the tool, an implicit equilibrium cycle search is accelerated by a modal expansion perturbation method that expands arbitrary flux perturbations on a large basis of λ-eigenmode harmonics. The harmonics are computed only at the reference state using Krylov subspace iterative methods, and substantial perturbations from this state are shown to be well approximated by computationally efficient algebraic expressions. The modal expansion method is coupled to the equilibrium method to produce the later-in-time response of each design perturbation, resulting in an explicit perturbation-accelerated equilibrium cycle method.Because the method determines the perturbed flux explicitly, a wide variety of core performance metrics may be tracked within optimization frameworks, including the performance of thermal hydraulics, fuel, economics, core mechanical, and transients. This capability strongly differentiates the method from traditional generalized perturbation theory approaches. The motivating end-use of the method is to evaluate objective functions in multidisciplinary optimization of advanced reactor designs, though many other applications are envisioned.

The Neutron Time-of-Flight Cross Section Program at the University of Kentucky - Adventures in Analysis II

Elastic and inelastic neutron differential cross sections are measured at the University of Kentucky Accelerator Laboratory (www.pa.uky.edu/accelerator/) at incident energies in the fast neutron region. The labo- ratorys facilities and instrumentation will be described and our measurement and analysis procedures outlined. Many corrections are required for neutron scattering experiments and the analysis utilizes information from many other cross section data sets and model calculations. Exploring and understanding the limitations of the foundational information and procedures are important for controlling the accuracy of the cross section results. We are examining the limitations in neutron detection efficiency, the normalization of (n,n � γ) cross sections, background reduction, spectrum stripping techniques, and attenuation and multiple scattering corrections. The resulting differential cross sections provide information on the compound elastic and coupled channels reaction mechanisms important for advanced reactor designs.

Joining of Oxide Dispersion Strengthened Steels for Advanced Reactors

The design, manufacture, and experimental analysis of structural materials capable of operation in the high temperatures, corrosive environments, and radiation damage spectra of future reactor designs remain one of the key pacing items for advanced reactor designs. The most promising candidate structural materials are vanadium-based refractory alloys, silicon carbide composites and oxide dispersion strengthened steels. Of these, oxide dispersion strengthened steels are a likely near-term candidate to meet required demands. This paper reviews different variants of oxide dispersion strengthened steels and discusses their capability with regard to high-temperature strength, corrosion resistance, and radiation damage resistance. Additionally, joining of oxide dispersion strengthened steels, which has been cited as a limiting factor preventing their use, is addressed and reviewed. Specifically, friction stir welding of these steels is reviewed as a promising joining method for oxide dispersion strengthened steels.

MATERIAL ISSUES FOR CURRENT AND ADVANCED NUCLEAR REACTOR DESIGNS

In all engineering applications, design and materials together determine the functionality and reliability of a device. This is particularly important in nuclear systems where the materials are pushed to their limits and phenomena not present anywhere else occur. In nuclear systems a combination of high temperature and pressure, stress, corrosive environment and high radiation environment combined causes significant materials challenges. Majority of commercial LWRs today are licensed for 40 years of operation, but many of them undergo lifetime extension to 60 or possibly 80 years. Materials degradation has always been a significant issue. However, due to the lifetime plant extension, finding materials that could sustain prolonged exposure to these extreme conditions has become a significant problem. In addition to the materials challenges in current LWRs, advanced reactors usually deal with even more difficult issues due to their operational requirements. Unusual heat transport media, such as liquid metals, liquid salts or other types of coolants, lead to a whole new set of material challenges. While corrosion has been the main issue, much higher operating temperatures create additional difficulties. In this paper, we present an overview of materials issues for current and advanced nuclear reactor designs.

The Laputa code for lattice physics analyses

A new lattice physics code called Laputa is currently under development at Tsinghua University. Aiming at the neutronics design of typical LWRs and advanced reactors, both the recent and traditional methodologies are integrated in the code. Laputa is able to model 2D arbitrary geometries by means of the combinational-geometry approach that has been commonly used in Monte Carlo codes. The method of characteristics (MOC), with the CMFD acceleration and OpenMP parallelism technologies, is implemented as the transport solver. The newly-proposed embedded self-shielding method (ESSM) is realized in Laputa for resonance treatment. Besides, an advanced matrix exponential method is implemented for solving the Bateman equations in the depletion module. Numerical validations against Monte Carlo solutions demonstrate that Laputa is very promising for lattice physics analyses.

AP1000® SBLOCA simulations with TRACE code

The AP1000® is an advanced pressurized water reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The performing of such systems must be evaluated through the performance of experiments and simulations with a variety of thermal–hydraulic codes. This paper presents the results which has been obtained for different SBLOCA break sizes with the best estimate TRACE V5.0 patch 2 thermal–hydraulic code and their comparison with those obtained by Westinghouse with NOTRUMP code. The main results show that TRACE code predicts a similar trend in all sequences with some differences that are expected to be an issue of the more conservative models and hypothesis assumed in the SBLOCA licensing analysis performed with NOTRUMP. Some particular characteristics of this reactor are also shown in this paper such as the importance of core makeup tank (CMT) and passive residual heat removal (PRHR) system for core cooling in case of small break sizes and the behavior of the plant in case of availability of half of the total passive safety injection systems which is the case of the double-ended direct vessel injection line break (DEDVI).

Dry-Cooled Supercritical CO2 Power for Advanced Nuclear Reactors

Abstract Currently, waste heat rejection from electrical power systems accounts for the largest fraction of water withdrawals from the U.S. fresh water table. Siting of nuclear power plants is limited to areas with access to a large natural supply of fresh or sea water. Due to a rise in energy needs and increased concern over environmental impact, dry air cooling systems are poised to play a large role in the future energy economy. In practice, the implementation of dry air-cooled condensing systems at steam plants has proven to be capital-intensive and requires the power cycle to take a significant efficiency penalty. These shortcomings are fundamental to dry-air steam condensation, which must occur at a fixed temperature. Closed-cycle gas turbines are an alternative to the conventional steam Rankine plant that allows for much improved dry heat rejection compatibility. Recent research into advanced nuclear energy systems has identified the supercritical CO2 (s-CO2) Brayton cycle in particular as a viable candidate for many proposed reactor types. The s-CO2 Brayton cycle can maintain superior thermal efficiency over a wide range of ambient temperatures, making these power systems ideally suited for dry air cooling, even in warm climates. For a sodium fast reactor (SFR) operating at 550 °C, thermal efficiency is calculated to be 43% with a 50 °C compressor inlet temperature. This is achieved by raising CO2 compressor inlet pressure in response to rising ambient temperatures. Preliminary design studies have shown that s-CO2 power cycle hardware will be compact and therefore well-matched to near-term and advanced integral small modular reactor (SMR) designs. These advantages also extend to the cooling plant, where it is estimated that dry cooling towers for an SFR-coupled s-CO2 power cycle will be similar in cost and scale to the evaporative cooling tower for a light-water reactor (LWR). The projected benefits of the s-CO2 power cycle coupled to dry air heat rejection may enable the long-awaited rise of next-generation nuclear energy systems, while redrawing the map for siting of small and large nuclear energy systems.

Resolution of Fission and Fusion Technology Integration Issues: An Upgraded Design Concept for the Subcritical Advanced Burner Reactor

The conceptual design of the subcritical advanced burner reactor (SABR), a 3000-MW(thermal) annular, modular sodium pool–type fast reactor, fueled by metallic transuranic (TRU) fuel processed from discharged light water reactor fuel and driven by a tokamak D-T fusion neutron source based on ITER physics and technology, has been substantially upgraded. Several issues related to the integration of fission and fusion technologies have been addressed, e.g., refueling a modular sodium pool reactor located within the magnetic coil configuration of a tokamak, achieving long-burn quasi-steady-state plasma operation, access for heating and current drive power transmission to a toroidal plasma surrounded by a sodium pool fast reactor, suppression of magnetohydrodynamic effects in a liquid metal coolant flowing in a magnetic field, tritium self-sufficiency in a TRU transmutation reactor, shielding the superconducting magnets from fusion and fission neutrons, etc. A design concept for a SABR that could be deployed within 25 years, based on the IFR/PRISM metal-fuel, sodium pool fast reactor technology and on the ITER fusion physics and technology, is presented. This design concept can be used for realistic fuel cycle, dynamic safety, and other performance analyses of a SABR.

Theory and implementation of nuclear safety system codes – Part II: System code closure relations, validation, and limitations

This is Part II of two articles describing the details of thermal-hydraulic system codes. In this second part of the article series, the system code closure relationships (used to model thermal and mechanical non-equilibrium and the coupling of the phases) for the governing equations are discussed and evaluated. These include several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models between the fields. These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models.Code verification and validation, including Separate Effects Tests (SETs) and Integral effects tests (IETs) is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities, but some of the more advanced reactor designs will push the limits of validation for the codes.Lastly, the limitations of the codes are discussed by considering next generation power plants, such as Small Modular Reactors (SMRs), analyzing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, innovative fuel types, liquid metal coolants, and many other design features that may not be fully analyzed by current system codes.This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations, validation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.