Commercial Power Reactors Research Articles

Evaluating Material Attractiveness of Minor Actinide Nuclear Fuel Intended for a Waste Transmutation Lead-Cooled Fast Reactor

Long-lived high-level waste from commercial nuclear power reactors is a problem that concerns stakeholders and scientists working in the back end of the nuclear fuel cycle. Nuclear waste transmutation is under investigation to tackle this problem, transforming nuclides that represent a long-term source of radioactivity, radiotoxicity, and heat into short-lived or stable nuclides. However, the transmutation process will require that several long-lived isotopes be separated from the spent nuclear fuel, which raises proliferation concerns. In this paper, we perform an investigation of the attractiveness characteristics related to the material used in a lead-cooled fast reactor system concept designed to burn minor actinides before and after irradiation. The materials evaluated are separated uranium, neptunium, plutonium, americium, and curium. We also evaluated grouped product materials, neptunium + americium and neptunium + plutonium. Additionally, we present potential safeguards and physical protection implications for the proposed materials. The main conclusion of this paper is that the separated neptunium and plutonium generated by the fast reactor are materials that deserve attention mainly related to physical protection measures.

Characterization and mapping of the neutron fields around Bruce Power’s 177Lu isotope production system

Bruce Power operates a first-of-its-kind isotope production system (IPS) that enables continuous production of 177Lu within Canada Deuterium Uranium (CANDU) commercial power reactors. Located on the reactivity mechanisms deck of Unit 7, just outside of reactor containment but in close proximity to the primary heat transport (PHT) pumps, this facility offers unique advantages for 177Lu production. However, employees working in this area encounter a radiation hazard which consists primarily of photoneutrons. These originate from the base of the PHT pumps and are only present when the reactor is operating. This study evaluates neutron exposure at Bruce Power’s IPS by using a nested neutron spectrometer (NNS) to determine the neutron energy spectra and absolute dosimetric quantities such as the ambient dose equivalent, H*(10). The results from the NNS are then compared to surveys performed by a portable neutron rem meter (Model NP-2 by Nuclear Research Corporation), routinely used by Bruce Power staff for workplace monitoring. While the Model NP-2 generally showed consistent results across locations, a 50% dose correction factor was identified when operators were harvesting 177Lu from the IPS. This finding highlights an opportunity to reduce the neutron dose that is assigned to operators when producing 177Lu.

Updates of the removable biological shielding blocks inside of the test cell, part of the test systems of IFMIF-DONES

The International Fusion Materials Irradiation Facility - DEMO Oriented Neutron Source (IFMIF-DONES) is a facility which is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of a neutron field having an neutron irradiation effect on materials like the one expected in a commercial fusion power reactor. The material specimens are irradiated in a containment structure named Test Cell (TC), which is part of the Test Systems (TS).The first protecting „barrier” against irradiation affecting the surrounding of the TC, are the Removable Biological Shielding Blocks (RBSBs). These ∼9 m height, ∼80 tons elements, which are formed by stainless steel liners filled with heavy concrete, need to be remotely handled as after the first experiments they will get dose rates above the hands-on limit. Irradiation will also result in a large nuclear heat power deposited in the shielding blocks, therefore needs to be actively water cooled by a system of embedded pipes to control the temperature.In this paper the updated design of the RBSBs is described, including the latest achievements and proposals for feasible manufacturing, lifting and positioning possibilities of the blocks inside of the TC respecting the given tolerances.

Practical methods for GPU-based whole-core Monte Carlo depletion calculation

Several practical methods for accelerating the depletion calculation in a GPU-based Monte Carlo (MC) code PRAGMA are presented including the multilevel spectral collapse method and the vectorized Chebyshev rational approximation method (CRAM). Since the generation of microscopic reaction rates for each nuclide needed for the construction of the depletion matrix of the Bateman equation requires either enormous memory access or tremendous physical memory, both of which are quite burdensome on GPUs, a new method called multilevel spectral collapse is proposed which combines two types of spectra to generate microscopic reaction rates: an ultrafine spectrum for an entire fuel pin and coarser spectra for each depletion region. Errors in reaction rates introduced by this method are mitigated by a hybrid usage of direct online reaction rate tallies for several important fissile nuclides. The linear system to appear in the solution process adopting the CRAM is solved by the Gauss-Seidel method which can be easily vectorized on GPUs. With the accelerated depletion methods, only about 10% of MC calculation time is consumed for depletion, so an accurate full core cycle depletion calculation for a commercial power reactor (BEAVRS) can be done in 16 h with 24 consumer-grade GPUs.

New strategies for a sustainable 99mTc supply to meet increasing medical demands: Promising solutions for current problems.

The continuing rapid expansion of 99mTc diagnostic agents always calls for scaling up 99mTc production to cover increasing clinical demand. Nevertheless, 99mTc availability depends mainly on the fission-produced 99Mo supply. This supply is seriously influenced during renewed emergency periods, such as the past 99Mo production crisis or the current COVID-19 pandemic. Consequently, these interruptions have promoted the need for 99mTc production through alternative strategies capable of providing clinical-grade 99mTc with high purity. In the light of this context, this review illustrates diverse production routes that either have commercially been used or new strategies that offer potential solutions to promote a rapid production growth of 99mTc. These techniques have been selected, highlighted, and evaluated to imply their impact on developing 99mTc production. Furthermore, their advantages and limitations, current situation, and long-term perspective were also discussed. It appears that, on the one hand, careful attention needs to be devoted to enhancing the 99Mo economy. It can be achieved by utilizing 98Mo neutron activation in commercial nuclear power reactors and using accelerator-based 99Mo production, especially the photonuclear transmutation strategy. On the other hand, more research efforts should be devoted to widening the utility of 99Mo/99mTc generators, which incorporate nanomaterial-based sorbents and promote their development, validation, and full automization in the near future. These strategies are expected to play a vital role in providing sufficient clinical-grade 99mTc, resulting in a reasonable cost per patient dose.

Analysis of diffusion coefficient correction methods applied for small-core, high-leakage reactors

Nodal diffusion codes have been successfully used for decades as a primary tool of commercial power reactor design, safety calculations and plant cycle simulations. The large-size, small-leakage property of these reactor cores and the appropriately generated and applied auxiliary parameters (such as albedos, discontinuity factors etc.) provide a calculation environment, where diffusion theory is fairly accurate, giving the industry the ultimate advantage of fast neutronic computation. Recently, several efforts have been made to extend this methodology to small-core, high-leakage research reactors, in which the validity of diffusion theory is not straightforward. In this paper, the appropriate generation of the diffusion coefficients and their effect on the overall performance of the simulations are investigated in light of recent developments. For the numerical analysis, the two-dimensional DIMPLE benchmarks and the BME TR core benchmark problems were chosen. Group constants were generated with the Serpent 2 Monte Carlo code, while nodal diffusion calculations were carried out with the PARCS code. The results were assessed in terms of multiplication factor, assembly level power and two-group flux distributions.

Ageing management review for reactor license renewal

Abstract The Nuclear Regulatory Commission (NRC) issues licenses for commercial power reactors to operate for up to 40 years and allows these licenses to be renewed for additional periods of 20 years. In 1995 the Commission issued Part 54 of Title 10 of the Code of Federal Regulations (10 CFR Part 54) specifying the regulatory requirements for license renewal. The license renewal rule focuses on managing the adverse effects of ageing of plant structures and components. The NRC staff has developed license renewal rule implementation guidance documents. The NRC staff uses a systematic approach in evaluating the adequacies of ageing management programs. The NRC has approved several license renewal applications to date and will likely be asked to approve dozens more over the next 10 to 15 years.

Assessment of the Generalization Ability of the ASTM E900-15 Embrittlement Trend Curve by Means of Monte Carlo Cross-Validation

The standard ASTM E900-15 provides an analytical expression to determine the transition temperature shift exhibited by Charpy V-notch data at 41-J for irradiated pressure vessel materials as a function of the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. The 26 free parameters included in this embrittlement correlation were fitted through maximum likelihood estimation using the PLOTTER—BASELINE database, which contains 1878 observations from commercial power reactors. The complexity of this model, derived from its high number of free parameters, invites a consideration of the possible existence of overfitting. The undeniable goal of a good predictive model is to generalize well from the training data that was used to fit its free parameters to new data from the problem domain. Overfitting takes place when a model, due to its high complexity, is able to learn not only the signal but also the noise in the training data to the extent that it negatively impacts the performance of the model on new data. This paper proposes the resampling method of Monte Carlo cross-validation to estimate the putative overfitting level of the ASTM E900-15 predictive model. This methodology is general and can be employed with any predictive model. After 5000 iterations of Monte Carlo cross-validation, large training and test datasets (7,035,000 and 2,355,000 instances, respectively) were obtained and compared to measure the amount of overfitting. A slightly lower prediction capacity was observed in the test set, both in terms of R2 (0.871 vs. 0.877 in the train set) and the RMSE (13.53 °C vs. 13.22 °C in the train set). Besides, strong statistically significant differences, which contrast with the subtle differences observed in R2 and RMSE, were obtained both between the means and the variances of the training and test sets. This result, which may seem paradoxical, can be properly interpreted from a correct understanding of the meaning of the p-value in practical terms. In conclusion, the ASTM E900-15 embrittlement trend curve possess good generalization ability and experiences a limited amount of overfitting.

A Preliminary Proposal for a Hybrid Lattice Confinement Fusion-Fission Reactor for Mobile Nuclear Power Plants

Scientists detected 2.45-MeV neutrons and in smaller yields 4- and 5-MeV neutrons in deuterated metals under a 2.9-MeV electron beam. Such discovery could allow the use of deuterated metals at temperatures below their melting point to provide nuclear fusion reactions. Such reactions could provide fast neutrons and energy in the form of heat. This work analyzed the results of some experiments to infer the neutron multiplication rate in such environments. It also considered the possible roles that such phenomena could play in a commercial nuclear power reactor under economic and compactness constraints. It seems the best way to promote nuclear fusion is the irradiation of deuterated metals by fast neutrons. This work presents the concept of a hybrid fission-fusion reactor using fissile or fertile fuel to generate heat and fast neutrons along deuterated metals providing excess neutrons (reactivity boost). Additionally, deuterated metals also may have a role in neutron moderation requiring less volume than other moderators (water or graphite). Such a reactor, given its reactivity boost, may burn radioactive residuals (transmutation) at affordable costs while generating power. Alternatively, this hybrid fission-fusion concept could also breed fissile fuel from fertile isotopes using natural uranium as seed.

Determination of oxidation rates and volatile oxidation products for HTGR graphite matrix material exposed to steam atmospheres

High-temperature gas-cooled reactors (HTGRs) in operation use tristructural isotropic (TRISO) particles embedded in graphite and carbonized resin matrix to form the fuel element. This graphite matrix material serves as a supportive structural element, heat transfer medium, and neutron moderator. In HTGR designs, fuel compacts are exposed to helium coolant, which facilitates high outlet temperatures (750°C<T<950°C) and subsequently greater thermal efficiency than current commercial power reactors. However, data pertaining to the oxidation resistance of graphite matrix material in HTGR accident or off-normal conditions (<50k kPa pH2O, 800°C<T<1600°C) is limited. In this study, both the oxidation behavior of matrix graphite material and its gas-phase products, including CO, CO2, and H2, are quantified in varied oxidant atmospheres using a coupled thermogravimetric analyzer and mass spectrometer. Oxidation rates reported here for varied steam (H2O [g]) atmospheres are predominantly linear and comparable with literature values in the range of tested temperatures (800–1200°C). Changes in dominant matrix oxidation products from primarily CO to a mixture of CO, CO2, and H2 were observed at higher temperatures (≥1000°C) and steam atmospheres (≥5 kPa pH2O). Kinetic data indicates that there was no shift in oxidation regime with chemical oxidation occurring at all temperatures and H2O (g) atmospheres tested. These data provide insight into the oxidation behavior of graphite matrix material and will inform future testing conditions, notably mixed atmospheric conditions, of HTGR fuel elements.

Failing the formative phase: The global diffusion of nuclear power is limited by national markets

Understanding the role of technology characteristics and the context in the diffusion of new energy technologies is important for assessing feasibility of climate mitigation. We examine the historical adoption of nuclear power as a case of a complex large scale energy technology. We conduct an event history analysis of grid connections of first sizable commercial nuclear power reactors in 79 countries between 1950 and 2018. We show that the introduction of nuclear power can largely be explained by contextual variables such as the proximity of a country to a major technology supplier (‘ease of diffusion’), the size of the economy, electricity demand growth, and energy import dependence (‘market attractiveness’). The lack of nuclear newcomers in the early 1990s can be explained by the lack of countries with high growth in electricity demand and sufficient capacities to build their first nuclear power plant, either on their own or with international help. We also find that nuclear accidents, the pursuit of nuclear weapons, and the advances made in competing technologies played only a minor role in nuclear technology failing to be established in more countries. Our analysis improves understanding of the feasibility of introducing contested and expensive technologies in a heterogenous world with motivations and capacities that differ across countries and by a patchwork of international relations. While countries with high state capacity or support from a major technology supplier are capable of introducing large-scale technologies quickly, technology diffusion to other regions might undergo significant delays due to lower motivations and capacities.

Development of high-fidelity neutron transport code STREAM

This paper presents a methodology developed and implemented in the neutron transport code STREAM to perform high-fidelity, multicycle, multiphysics simulations of light water reactor whole-core problems. STREAM uses state-of-the-art methodologies to achieve high accuracy and computational performance. Further, it can compute fine-mesh neutron transport solutions for three-dimensional (3D) whole-core problems using the 3D method of characteristics/diamond-difference (MOC/DD). For multigroup calculation, STREAM generates problem-dependent multigroup cross-sections using intrinsic cross-section libraries and a resonance self-shielding method. To perform the multicycle analysis of commercial power reactors, many advanced methods and features have been developed and implemented. In this paper, the methodologies and features of STREAM are presented, followed by the whole-core simulation capability.

Skyshine Calculations for a Large Spent Nuclear Fuel Storage Facility with SCALE 6.2.3

The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous applications, including dose rate analyses of complex consolidated interim storage facilities (CISFs). A licensed CISF could be used to store tens of thousands of tonnes of spent nuclear fuel discharged from commercial power reactors using various cask and storage pad designs. A CISF design must comply with the regulatory requirements provided in 10 CFR Part 72, including requirements related to annual dose limits applicable to real individuals located beyond the area controlled by the licensee. Therefore, calculating a dose to the public is a necessary part of the licensing process for the construction of a CISF. These calculations are very challenging because of the complexity of the CISF design and the low magnitude of dose rate at large distances from the facility. This paper describes detailed far-field dose rate calculations performed for a proposed CISF using MAVRIC, the Monte Carlo radiation shielding sequence in SCALE 6.2.3, with automated variance reduction based on discrete ordinates calculations. The method presented in this paper uses a detailed Monte Carlo radiation transport simulation in one step from source to dose rate. A series of independent simulations was made using the complete site geometry (all casks present), but with only one cask containing radiation sources to obtain the dose rate maps produced by each storage cask. The CISF dose rate map was obtained by adding the dose rate maps produced by the independent individual cask simulations. Ample volumes of air and soil extending beyond the location of interest for dose rate calculation were included in the calculation model to properly simulate important radiation attenuation and scattering events that affect far-field dose rates. A comprehensive sensitivity study is included in this paper to illustrate the importance of selecting appropriate air volume, mass density, and composition for CISF skyshine dose rate calculations. Dry soil and soil containing water were analyzed to determine their effects on groundshine radiation.

A Study on Reusable Metal Component as Burnable Absorber Through Monte Carlo Depletion Analysis

After nuclear power plants are permanently shut down and decommissioned, the remaining irradiated metal components such as stainless steel, carbon steel, and Inconel can be used as neutron absorber. This study investigates the possibility of reusing these metal components as neutron absorber materials, that is burnable poison. The absorption cross section of the irradiated metals did not lose their chemical properties and performance even if they were irradiated over 40-50 years in the NPPs. To examine the absorption capability of the waste metals, the lattice calculations of WH 17×17 fuel assembly were analyzed. From the results, Inconel-718 significantly hold-down fuel assembly excess reactivity compared to stainless steel 304 and carbon steel because Inconel-718 contains a small amount of boron nuclide. From the results, a 20wt% impurity of boron in irradiated Inconel-718 enhances the excess reactivity suppression. The application of irradiated Inconel-718 as a burnable absorber for SMR core was investigated. The irradiated Inconel-718 impurity with 20wt% of boron content can maintain and suppress the whole core reactivity. We emphasize that the irradiated metal components can be used as burnable absorber materials to control the reactivity of commercial reactor power and small modular reactors.

Assembly design of a fluoride salt-cooled high temperature commercial-scale reactor: Neutronics evaluation and parametric analysis

In this study, a novel assembly design is proposed for a fluoride salt-cooled high-temperature commercial-scale (FHCR) reactor. It employs tristructural isotropic (TRISO) fuel particles nestled within removable cylindrical beryllium carbide (Be2C) moderator blocks, which are further contained within prismatic graphite blocks. As the name implies, the FHCR is a preconception of a 3400 MW(t) commercial power reactor that uses FLiBe (LiF-BeF2) as the primary coolant of choice. This paper uses the SERPENT 2 code to conduct a parametric neutronics study on the two-dimensional lattice assembly design of the FHCR. The calculations examine the effects of various fuel enrichment levels, TRISO particle packing fractions, fuel compacts’ pitch sizes, and moderating materials on the cycle length, while also determining the neutron spectrum and various nuclide inventories. Finally, a preliminary core is modeled based on the study conducted on the assembly design. Based on the negative values of the fuel temperature coefficients (FTC), moderator temperature coefficients (MTC), and coolant temperature coefficients (CTC) obtained, the design is determined to be safe. This new assembly design is also able to achieve keff>1 for approximately 3.55 years, translating to a burn-up of 160.6 MWd/KgU.

Microscale Isotopic Variation in Uranium Fuel Pellets with Implications for Nuclear Forensics.

Until recently, the analysis and identification of nuclear fuel pellets in the context of a nuclear forensics investigation have been mainly focused on macroscopic characteristics, such as fuel pellet dimensions, uranium enrichment, and other reactor-specific features. Here, we report microscale isotopic heterogeneity observed in different fuel pellet fragments that were characterized in situ by nanoscale secondary ion mass spectrometry (NanoSIMS). The materials analyzed include fuel fragments obtained as part of the Collaborative Materials Exercise (CMX-4) organized by the Nuclear Forensics International Technical Working Group (ITWG), as well as a fuel pellet fragment from a commercial power reactor. Although the commercial fuel pellet showed a homogeneous 235U/238U ratio across the sample (within analytical error), NanoSIMS imaging of the CMX-4 fuel pellet fragments showed distinct microscale variations in the uranium isotopic composition. The average 235U enrichments were 2.2 and 2.9% for the two samples; however, the measured 235U/238U ratios varied between 0.0081 and 0.035 (0.79-3.3 atom % 235U) and between 0.0090 and 0.045 (0.89-4.3 atom % 235U). The measurement of 236U in one of the CMX-4 samples suggested the use of at least three uranium oxide powders of different isotopic compositions ("source terms") in the production of the pellets. These variations were not detected using the conventional bulk, macroscopic techniques applied to these materials. Our study highlights the importance of characterizing samples on the microscale for heterogeneities that would otherwise be overlooked and demonstrates the potential use of NanoSIMS in guiding further nuclear forensic analysis.

Investigative Study of Neutronic Safety Parameters of HPR and EPR Using the MCNP Code

As part of technology assessment of proposed commercial nuclear power reactor technologies for Ghana’s Nuclear Power Programme, the neutronic safety parameters of the European Pressurized Reactor (EPR) and High Temperature Pressurized Reactor (HPR) reactor technologies are theoretically analyzed and compared. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality, and neutron behavior at various operating conditions. The HPR, which is still under construction and under theoretical safety analysis, showed good inherent safety features comparable to the already existing EPR technology.

Evaluating Reactor Antineutrino Signals for WATCHMAN

Increasing the distance from which an antineutrino detector is capable of monitoring the operation of a registered reactor, or discovering a clandestine reactor, strengthens the Non-Proliferation of Nuclear Weapons Treaty. This paper presents calculations of reactor antineutrino interactions from quasi-elastic neutrino-proton scattering and elastic neutrino-electron scattering in a water-based detector operated ≳ 10 km from a commercial power reactor. It separately calculates signal from the proximal reactor and background from all other registered reactors. The main results are differential and integral interaction rates from the quasi-elastic and elastic processes. There are two underground facilities capable of hosting a detector (∼ 1 kT H2O) project nearby (L ∼ 20 km) an operating commercial reactor (Pth ∼ 3 GW). These reactor-site combinations, which are under consideration for project WATCHMAN, are Perry-Morton on the southern shore of Lake Erie in the United States and Hartlepool-Boulby on the western shore of the North Sea in England. The signal rate from the proximal reactor is about five times greater at the Morton site than at the Boulby site due to shorter reactor-site separation distance, larger reactor thermal power, and greater neutrino oscillation survival probability. Although the background rate from all other reactors is larger at Morton than at Boulby, it is a smaller fraction of the signal rate from the proximal reactor at Morton than at Boulby. Moreover, the Hartlepool power plant has two cores whereas the Perry plant has a single core. The Boulby site, therefore, offers an opportunity for remotely monitoring the on/off cycle of a reactor core under more stringent conditions than does the Morton site.

The Significance of Carbon 14 in Graphite Reactor Components at End of Generation

It is estimated that there are at least 250,000 tonnes of irradiated graphite worldwide that will require eventual disposal. This graphite arises from a number of sources, but principally comprises moderator and structural materials for experimental reactors, production reactors, commercial power reactors and fuel assemblies. In the UK, a significant proportion of its irradiated graphite is classified as Intermediate Level Waste. Such waste is not heat generating but has a radioactive content exceeding 4 GBq per tonne alpha or 12 GBq per tonne beta/gamma activity. While the classification of waste is not consistent across states and proposals by individual states for the management of their graphite waste vary considerably, a common interest is the nature and distribution of its radioactive content. The radionuclides in irradiated graphite presenting the most significant long-term hazard are Carbon 14 (C-14) and Chlorine 36 (Cl-36) with half-lives of 5730 and 301,000 years respectively. For a better understanding of the way in which C-14 is produced, its distribution within irradiated graphite and realistic quantification of activity can potentially lead to improved characterization to validate its status within current or future waste classifications, segregation to reduce Intermediate Level Waste volumes, or treatment to reduce activity enabling re-classification as Low Level Waste. This paper reviews all these issues and then focuses on the significance of C-14. Some findings from a National Nuclear Laboratory study of C-14 levels in carbonaceous deposits and the underlying Magnox reactor graphite are presented to illustrate the need for thorough characterization of the waste material. These results are discussed in the context of aqueous leaching of C-14 from irradiated graphite and potential treatment options to minimize aqueous release. The paper concludes with some broader observations on the significance of C-14 in nuclear reactor graphite components and how these issues should be considered when preparing the lifetime management of new nuclear plant.

Preliminary Evaluation of Decommissioning Wastes for the First Commercial Nuclear Power Reactor in South Korea

Preliminary Evaluation of Decommissioning Wastes for the First Commercial Nuclear Power Reactor in South Korea