Subcritical Reactor Research Articles

Thermo-mechanical behaviour of the MYRRHA core support structure

This work is part of the research on the design of MYRRHA primary system. MYRRHA, a multipurpose research facility, being developed at SCK-CEN, is based on the Accelerator Driven System (ADS) concept. This research facility intends to show the applicability of the ADS principle, and aims to demonstrate, among other objectives, the feasibility of efficient transmutation of high-level nuclear waste. The present study investigates, mainly, the thermo-mechanical behaviour of the core support bundle, fuel assemblies and core devices, in particular, the deformation in terms of the displacement and gaps distribution. To reach this goal, which is arduous, an efficient modeling methodology based on the finite element method has been devised. An effective solution strategy has been formulated to carry out a highly challenging analysis due to severe numerical issues and instabilities, making the solutions particularly difficult to reach for some cases. Analysis cases are differentiated by some friction coefficient values; and the presence or not of a spring pre-compression force. Despite serious convergence issues, converged solutions have been achieved for all the cases studied. From the results analysis, a remarkable feature observed is the stability of the solutions obtained. And, on the whole, the value of the friction coefficient used does not change significantly the displacement, the gaps patterns, and the stresses. Similarly, account of the pre-compression force does not affect much the global behaviour. The kinematic quantities found should be useful for some preliminary reactivity predictions of the reactor.

Online identification of trips caused by the external proton source in an ADS reactor

New fields in the domain of signal analysis began with the improvement of the techniques that gave rise to Wavelet Transforms in the late 1980s and early 1990s. Another important element that allowed the development and use of new methods in signal analysis was the increase in processing power of modern computers, which also contributed to the development and application of Wavelet Transforms. This work uses the Haar Wavelet in the analysis of the neutron flux of an Accelerator Driven System (ADS) reactor to provide a method for the online identification of trips in the external neutron source as produced by transients. In order to obtain the necessary data, an ADS reactor was simulated, based on the MYRRHA reactor, with the use of the Serpent Physics reactor code.

Effect of Irradiation on Corrosion Behavior of 316L Steel in Lead-Bismuth Eutectic with Different Oxygen Concentrations

In an accelerator-driven system (ADS), the beam window material of the spallation neutron target is heavily irradiated under severe conditions, in which the radiation damage and corrosion co-occur because of high-energy neutron and/or proton irradiation in the lead–bismuth flow. The materials used in ADSs must be compatible with the liquid metal (lead–bismuth eutectic (LBE)) to prevent issues such as liquid metal embrittlement (LME) and liquid metal corrosion (LMC). This study considers the LMC behavior after ion irradiation of 316L austenitic steel for self-ion irradiations followed by the corrosion tests in LBE with critical oxygen concentration. The 316L samples were irradiated by 10.5 MeV-Fe3+ ions at a temperature of 450 °C, up to 50 displacements per atom (dpa). After the corrosion test performed at 450 °C in LBE with low oxygen concentration, a surface of the nonirradiated area was not oxidized but appeared with locally corrosive morphology, Ni depletion, whereas an iron/chromium oxide layer fully covered the irradiated area. In the case of the corrosion surface with high oxygen concentration in LBE, the surface of the nonirradiated area was covered by an iron oxide layer only, whereas the irradiated area was covered by the duplex layers comprising iron and iron/chromium oxides. It is suggested that irradiation can enhance the oxide layer formation because of the enhancement of Fe and/or oxygen diffusion induced by the radiation defects in 316L steel.

Arbitrarily Large Neutron Amplification in Subcritical Nuclear Reactors

In a subcritical reactor, each neutron produces only keff < 1 neutron per generation (asymptotically and on average), and thus the neutron population decreases exponentially in the absence of external source. The chain reaction is thus easy to stop, making such reactors inherently stable. Interestingly, and contrary to common wisdom, there is no relation between keff, the external source intensity, and the output power of the reactor. In this paper, I present various possible strategies to exploit this fact, and apply them to the design of a rudimentary multi-layer system that allows to reach an arbitrarily large number of fissions per source neutron, while keeping keff < 1 fixed (I used keff=0.97). The behavior of the amplifier is verified with simple Monte-Carlo transport simulations. The proposal admittedly brings complications and subtleties that need to be studied further, but, if developed, could allow to drive subcritical reactors with faint neutron sources, e.g. radioisotope based.

Simulation of transients in ads reactors in three-dimensional geometry

Accelerator-Driven System, ADS, belong to the new generation of advanced reactors being developed that promise to drastically reduce the life of radioactive waste by, for example, the transmutation process. Subcritical reactor designs of the ADS type have attracted worldwide attention and are the subject of research and development in several countries. The purpose of this work is to simulate transients associated with ADS. It adopted the neutron diffusion model that leads the spatial kinetics equations. These equations are solved by the known numerical method of finite differences. The simulations are performed considering transients related to the variations in the intensity of the proton flux provided by the particle accelerator acting in a sub-critical reactor in three-dimensional geometry for two energy groups and six groups of delayed neutron precursors.

Techno-economic assessment of subcritical water hydrolysis process for sugars production from brewer’s spent grains

Breweries generate a high amount of brewer’s spent grains (BSG), which is a valuable feedstock for industrial applications. In the current study, a techno-economic assessment of the flow-through subcritical water hydrolysis reactor to produce sugars from BSG was performed. Simulations were done for a sequential hydrolysis process in three-extractor vessels of 10 L (pilot-plant) and 500 L (industrial-plant), coupled or not to a separation system. The sugar separation system was composed of a five-zone simulated moving bed (SMB) process. A study on the cost of manufacturing (COM), profitability indicators, and sensitivity analysis was conducted to verify the project feasibility. Moreover, the mass and energy balance of the industrial process was accomplished to evidence the main operational parameters. Simulation results indicated that the scale-up process from pilot to industrial scale reduced the COM by approximately 80 %. In the process coupled with the SMB separation system, arabinose and galactose represented 83.68 % of the costs for sugars separation. Arabinose COM decreased from 64.10 USD kg–1 in pilot-plant to 7.22 USD kg–1 in industrial-plant. The implementation of a separation system recovering six sugars with high added-value can be an advantage in the industrial-plant process when comparing to the process without SMB process, which produces a single hydrolysate fraction with low commercial value. Finally, the integrated subcritical water hydrolysis of BSG coupled with a separation system can be a promising alternative to produce different concentrated sugars in a biorefinery concept.

Estimation of fission yield and cross-section for thorium-based molten salt reactor and accelerator driven system

Thorium is the primary candidate fuel in Thorium-based Molten Salt Reactor (TMSR) and Accelerator Driven System (ADS). However, to utilize Thorium in the reactor and accelerator systems, nuclear data in the literature are not sufficient for neutron-induced fission reaction process. Therefore, we estimated neutron fission cross-section [Formula: see text] (mb), independent [Formula: see text] and primary fission fragment yield [Formula: see text], mass-dependent average neutron multiplicity [Formula: see text], energy-dependent average neutron multiplicity [Formula: see text] and neutron emission multiplicity distribution [Formula: see text] of [Formula: see text] for different neutron incident energy values, 1 MeV, 1.5 MeV, 3 MeV, 5.5 MeV, 7.7 MeV, 10 MeV, 14.8 MeV and 20 MeV, where the energy region is important for the reactor and accelerator systems used in the industry. The obtained results are compared with experimental data obtained from the literature and are discussed for each energy value.

Toxicity Reduction with Total Volume Control in Nuclear Waste

When total volume control is applied to toxicity in nuclear waste management it becomes a limiting factor for the permittable total operating capacity of nuclear reactors. An alternative conceptual scenario is proposed, aimed at toxicity reduction through partitioning and transmutation. Specifically, the electricity generation capacity could be increased by transmutation of 90Sr and 137Cs. Simultaneously, the cooling time before disposal is reduced to 50 years from the 300 years required by existing scenarios, such as the accelerator-driven system. Finally, the scenario is also found to be feasible in terms of energy balance and cost, using an Li(d,xn) reaction neutron source with a deuteron accelerator for transmutation.

Reactivity determination in a subcritical reactor: Computational, analytical and experimental methods

In this paper, an analysis of reactivity determination methods at Jordan Subcritical Assembly (JSA), which is driven by an external neutron source, is presented. Results from previous computational analysis estimate the effective multiplication factor, keff, to be 0.96553 ± 0.00004. Calculations of reactivity in a subcritical system predicts the source multiplication factor, ks, to be 0.9637 ± 0.0007, which is very close to the computed value of keff. The experimental values of the prompt neutron decay constant (α) using Rossi-α and Feynman-α methods, were 551.5 ± 8.0 (s−1) and 555.5 ± 9.8 (s−1), respectively. Using the values for effective delayed neutron fraction, βeff, and the mean neutron generation time, Λ, from the computational analysis, give estimations for keff by Rossi-α and Feynman-α to be 0.9548 ± 0.0009 and 0.9544 ± 0.0011, respectively, which are lower than the computed value by 1.1%.

Preliminary study on spallation target characteristics in an Accelerator-Driven system

This study aims to investigate the neutronic characteristics of spallation targets for Accelerator-Driven subcritical System (ADS) and find the optimal target design for reducing the strength of the required beam current. All the calculations were conducted using the MCNP6.1 with cross-section library ENDF/B-VII. In this study, the influence of several parameters on spallation targets is investigated, such as neutron production with various spallation target layouts, spallation neutron distribution with different proton beam energy levels, and spallation neutron spectrum.

Bubble Behaviors in a Fuel Assembly of Accelerator Driven System

The accelerator driven system (ADS) is a nuclear reactor that can burn spent nuclear fuel with inherent safety features and operational flexibility. China is developing an ADS research device named China Initiative Accelerator Driven System (CiADS). The steam generator is one of the most important parts in the primary circuit of CiADS. The rupture of steam generator tubes may cause a significant impact on the reactor core due to potential risks of steam bubble intrusion into the core region. When the steam generator tube-rupture accident occurs, the primary coolant turns to a two-phase flow, which may cause variations of flow conditions through fuel assemblies. This paper adopted the open-source computational fluid dynamics calculation software OpenFOAM, and performed numerical simulations of a two-phase flow scenario in the ADS fuel assembly with the volume-of-fluid method. By changing boundary conditions at the assembly inlet, different scenarios of steam bubble intrusion were simulated and compared, based on our preliminarily conclusion that steam bubbles evenly distributed at the assembly inlet tend to accumulate at the periphery region, which partly confirmed the necessity of building experimental facilities for relevant two-phase flow research.

Applicability of several Feynman-α formulae to a subcritical thermal reactor

ABSTRACT For an university training and research reactor of Kindai University, the Feynman-α analyses based on several formulae are carried out to examine the applicability of these formulae to the subcritical thermal reactor system. The original formula, in which the effect of delayed neutrons is neglected, results in a prompt-neutron decay constant sensitive to an upper limit of the gate time range of the analysis and consequently has difficulty in determining the decay constant. This difficulty originates from the neglect of delayed neutrons. An improved formula, where a term proportional to the gate time is added to the original formula, gives a desirable decay constant insensitive to the upper limit and the insensitive decay constant well agrees with that obtained from the Rossi-α analysis. The application of a difference filtering technique also leads to a successful result similar to the above improved formula. An alternative index Z to the traditional index Y of the original formula has been sometimes employed as an indication of the non-Poisson nature of the counting statistics. The Z is defined as dividing the Y by the mean number of neutron counts accumulated in the gate time. The formula for the Z never results in a reasonable decay constant even though the effect of delayed neutrons is considered. This study suggests that the improved formula considering delayed neutrons for the Y and the formula for the difference filtering technique should be employed to determine the prompt-neutron decay constant of thermal reactors.

Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident.

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.

High heat removal technique for a lithium neutron generation target used for an accelerator-driven BNCT system

Accelerator-driven neutron sources have been developed worldwide for boron neutron capture therapy (BNCT) instead of nuclear reactors. Nagoya University is currently developing a compact DC accelerator driven BNCT system using a sealed lithium (Li) target. The target receives a high heat load by a proton beam, and the heat then requires efficient removal. This study developed a high-efficiency heat removal technique that uses V-shaped staggered ribs on the heat removal side of the flow channels. We performed experiments using an electron beam to confirm the technique performance. The results showed double the heat removal performance compared to the smooth surface channel. The value expected to sufficiently remove 6.6 MW/m2 of heat load provided by the proton beam at the Nagoya University neutron source for BNCT.

Benchmark on neutron flux spatial effects in subcritical system based on IRT-4 M fuel for near-core positions

The paper presents a benchmark aimed at assessing the accuracy of Monte Carlo codes to simulate the neutron detector response in near-core geometry of a subcritical water-moderated reactor. The experiment included six levels of subcriticality with keff from 0.93 to 0.996, and four fixed detector positions. Studied parameters included: first, the prediction of the change in detector response with the change in subcritical multiplication at a given detector position; second, the prediction of the change in neutron flux spatial distribution; third, the effectiveness of calculated correction factors to decrease the inconsistency in reactivity determined using detectors at various locations. Calculations were performed in MCNP6 code using two nuclear data sets. Both the general source and the eigenvalue (criticality) mode of calculation were studied. Finally, the effect of the offset between the calculated and experimentally determined critical condition on the prediction of detector response in the subcritical state was analysed.

Comparative Analysis of Reactivity Calculations for the CERMET Fueled ADS with Serpent and MCNP6 Codes

The ADS (accelerator driven system) is recognized as a promising system to annihilate the radioactivity of nuclear waste with its inherent safety feature and waste transmutation potential. Thus, conceptual designs of ADS are widely carrying out. In order to verify the accuracy of an innovative ADS core modeling by using simulation codes, the reactivity calculations of CERMET fueled ADS were conducted using two Monte Carlo codes, Serpent and MCNP6 with ENDF/B-VII.0 library. The comparison shows a good agreement between two codes including the eigenvalue (less than 50 pcm) and fuel temperature feedback (discrepancy is within the standard deviation). It implies that the ADS was modelled successfully and can be used for further investigation.&#x0D; Keywords: CERMET fueled ADS, Serpent code, MCNP6, reactivity calculation.

Interaction of Liquid Lead Bismuth and Materials in Spallation Target

Material choices for liquid lead bismuth spallation target are some of austenitic stainless steel, ferrite martensitic steel and cold-worked austenitic stainless steel. In order to ensure materials resistance to irradiation and corrosion as well as compatibility with lead bismuth, it is appropriate to lower the incident proton current density and the process temperature, in which temperature range engineering design can control to work, especially in ADS (Accelerator-Driven nuclear transmutation System) concept. The lower limit temperature is determined from the physical melting temperature and the engineering efficiency of the steam generator involved in process control. The material related issues for liquid lead bismuth are mass loss by impinging secondary flow, wettability at the device interface for ultrasonic waves application, detachable control of the slag in the flowing system, stabilized electrical resistance between the material and the liquid lead bismuth interface. Electromagnetic fluid analyses show how flow rate relates electrical resistivity of flow channel material.

Coupled multi-physics simulation for the evaluation of an accelerator-driven Aqueous Homogeneous Subcritical System for medical isotope production

The most common medical isotope (99mTc) is widely used for the diagnosis and therapy of complex diseases, with approximately 30–40 million nuclear medicine procedures worldwide every year. Although 99mTc can be directly produced in particle accelerators, it is generally obtained from the beta-decay of 99Mo. The utilization of an accelerator-driven Aqueous Homogeneous Subcritical System to produce 99Mo has several advantages and therefore, represents an attractive alternative to the conventional method of irradiating uranium solid targets in nuclear research reactors. This paper aims to perform a multi-physics evaluation of an Aqueous Homogeneous Subcritical System conceptual design with low enriched uranium (LEU), driven by a D-T neutron generator, to produce 99Mo in safe conditions. A coupled iterative calculation methodology was implemented to predict the steady state operation of the system at 20 kWth using the MCNP6 and ANSYS-CFX 18.1 codes for neutronic and thermal-hydraulic calculations, respectively. The main neutronic and thermal-hydraulic parameters were monitored to verify if they met the established convergence criterion of 1% to stop the calculation process. The methodology convergence was reached after four calculation steps when the most changing parameter, average fuel solution velocity, satisfied the criterion. However, others such as the average bubble velocity, the ion beam current, the fuel solution height and the effective multiplication factor varied less than 0.1% in the fourth step. A burnup calculation has shown that the system is capable of producing 114.3 six-days Ci of 99Mo in a one-week operation period and significant amounts 133Xe, 131I and 89Sr.

Uncertainty Quantification of Lead and Bismuth Sample Reactivity Worth at Kyoto University Critical Assembly

Uncertainty quantification of lead (Pb) and bismuth (Bi) sample reactivity worth is numerically determined using the SCALE6.2 code system and experimental results obtained from the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA) to demonstrate the sensitivity coefficients of aluminum (Al) and Bi scattering reactions. From the results of the numerical analyses, the impact of 27Al and 209Bi scattering cross sections obtained using SCALE6.2/TSAR is disclosed on the Bi sample reactivity worth using Al reference and Bi test samples, although the uncertainty itself is small in the Bi sample reactivity worth. In future studies, the actual impact of 209Bi inelastic scattering reactions in liquid Pb-Bi eutectic needs to be considered under numerical simulations of the void reactivity in the accelerator-driven system. Also, in the KUCA experiments, conventional modeling of void evolution by the Al reference sample is expected to be altered by a void (air) in terms of the 27Al elastic and inelastic scattering reactions of the Al reference sample.

High power particle accelerator for driving the nuclear waste transmutation system at nuclear power plant

The existing technology for treatment of NPP nuclear waste is a deep geological disposal technology. But this technology is still has possibility of leakage due to heating caused by radioactivity process from long lived elements of nuclear waste. The development of high power particle accelerator technology which can produce high power proton beam to induce nuclear spallation reaction has introduced a new accelerator-based technology for treatment of nuclear waste from NPP named the Accelerator-driven Transmutation Waste (ATW). This paper describes result of study on status of accelerator based transmutation system for NPP nuclear waste. ATW is subcritical nuclear reactor combined with high power proton accelerator for transmutation of nuclear waste utilizing neutrons from spallation reaction due to collision of high power proton beam with heavy metal material as spallation target. The types of accelerators used for ATW are cyclotron, linac and synchrotron which able to produce proton beam of 10 up to 100 MW beam power, 10 up to 100 mA beam current and 0.6 up to 1.5 GeV beam energy. Demonstration scale ATW requires 1 up to 2 MW of proton beam power to generate 50 up to 100 MWth of thermal power in subcritical reactor. Whereas industrial ATW requires 10 up to 75 MW of proton beam power to generate hundred MWth of thermal power in subcritical reactor. Several countries have roadmap of ADS research and development which the main goal of nuclear waste transmutation using ADS called ATW. ATW also can be used for electrical energy production which is safer and more reliable.