Accelerator-driven Subcritical Reactor Research Articles

Availability and reliability statistics at ISIS and at other high-power proton accelerators

Detailed availability statistics, and reliability statistics in terms of trip rates, between 1994 and 2021 at the ISIS Spallation Neutron Source are presented. The average of the availabilities for 121 irradiation campaigns spanning the better part of three decades is 86% with a standard deviation of 7%. Trip rates at ISIS are very similar to trip rates at high-power proton machines of different accelerator ‘architecture’ elsewhere, and altogether these trip rates can be fitted quite well over several orders of magnitude by the simple power law Ṅ=6t−0.75 where Ṅ is the number of trips per day longer than time t (t in minutes). Typical trip rates for present-day accelerator systems would appear to be too high to be suitable for acceptable operation of accelerator-driven subcritical reactor systems.

High-power Fixed-Field Accelerators

We report the state of the field of High-Power Fixed-Field Accelerators (with an emphasis on cyclotrons) as discussed by international experts during a three-day workshop of the same name in 2021. The workshop was part of the Snowmass'21 Community Planning Exercise. Here, we take stock of the world inventory of high-power fixed-field accelerators, assess available beam currents and beam powers, and investigate limitations. Furthermore, we evaluate the role of these machines in particle physics, directly used or as injectors to other machines, and in industry, as drivers for (medical) isotope production and, potentially, for accelerator-driven systems and sub-critical reactors. Finally, we discuss novel concepts and cutting-edge developments to push the available current higher at several energy scales, thereby increasing relative power. Highlights include new spiral inflector types, direct RFQ injection, H2 + acceleration, utilizing vortex motion, and self-extraction schemes. We also discuss modern computational frameworks to optimize accelerators more efficiently, and better describe the relevant physical processes in simulations.

The Influence Mechanism of Neutron Kinetics of the Accelerator-Driven Subcritical Reactor Based on the Fast/Thermal Neutron Spectra by Monte Carlo Homogenization Method

For the sake of understanding the mechanism of deep subcriticality and high heterogeneity of neutron fluence rate in time–space on the neutron kinetics of the Accelerator-driven Subcritical Reactor Subcritical Reactor (ADSR) under varied beam transients and neutron spectra. A Monte Carlo homogenization approach for the neutron time–space kinetics of the ADSR is proposed in this study, and the influence mechanism on the kinetic parameters of the ADSR under varied neutron spectra, subcriticality, and beam transients is examined. The results show that the Monte Carlo homogenization for the α eigenvalue mode is more adaptable to the subcriticality characteristics under varied subcriticality; under beam transients, the relative differences in the kinetic parameters of the different modes of the ADSR with fast/thermal spectra increase with the depth of subcriticality, and the differences in neutron generation time for varied modes are larger than those of effective fraction of delayed neutron. Thus, it is recommended to use a more adaptable Monte Carlo homogenization method for the time–space kinetics of ADSR, and the effects of the high heterogeneity of neutron fluence rate and deep subcriticality in time–space on the neutron generation time should be considered.

Effect of Higher-Order Harmonics on the Steady-State Neutron Flux in Accelerator-Driven Subcritical Reactors

Higher-order neutron fluxes (i.e., higher-order harmonics) are widely applied in perturbation theory and modal kinetics, and they are important for research on the physical characteristics of accelerator-driven subcritical reactors (ADSRs). This paper presents a computational scheme for reconstructing the neutron flux in the steady state according to the biorthogonal properties of the forward and adjoint neutron fluxes, which can be used to analyze how higher-order harmonics affect the steady-state neutron flux under λ - and prompt α -modes. Simulation results indicated that a modal synthesis method based on λ - and prompt α -modes can effectively reconstruct the steady-state neutron flux and core power in an ADSR with a power reconstruction accuracy of within 5%. The higher-order harmonics can be classified into three types according to their physical characteristics: the first type contributes significantly to the steady-state neutron flux, the second type contributes almost nothing to the steady-state neutron flux, and the third type contributes nothing to the steady-state neutron flux. The external neutron source contributes only to specific harmonic expansion orders, which are characterized by significant axial and radial symmetry for both the λ - and prompt α -modes.

The feasibility study of the Sub-criticalization of the Holos small modular reactor driven by an electron accelerator

Small modular reactors (SMRs) and accelerator-driven subcritical reactors (ADSs) have many potentials, which have attracted the attention of many researchers. The present study proposes the sub-criticalization of a small portable reactor to increase its safety. The operation of the Holos mobile power generator is examined at the subcritical level to produce cost-effective power. The UO2-fueled Holos subcritical core is modelled by the MCNPX2.6 code. The external photoneutron source was considered with a 100 MeV, 1 mA electron accelerator. The Holos simulated core is discussed in the subcritical level to evaluate the neutronic performance and safety parameters. Power distributions were compared inside the Holos simulated core in critical and subcritical levels. The carried-out calculations show that in addition to the subcritical level operation, negative reactivity coefficient in the Holos subcritical core ensures the core inherent safety. Results show that the neutron and photon flux magnitudes within the Holos core in the subcritical level are greater than the critical state.

Study of External Source Effects for Different Multibeam Concept for Accelerator-Driven Subcritical Reactor for Nuclear Waste Transmutation (ADS-NWT)

The multibeam concept for the accelerator-driven subcritical reactor (ADS) has advantages in the power distribution of the core, and it could reduce the requirement of proton beam current intensity for each beam. In comparison with the single-beam concept, the multibeam concept could also reduce the thermal load of the beam window significantly. This paper focuses on the study of external source effects for different multibeam concepts for an ADS for nuclear waste transmutation (ADS-NWT). The different multibeam concepts include the three-beam, four-beam, six-beam, and seven-beam concepts for the ADS-NWT. By using the calculation tools FLUKA and SuperMC with the nuclear data library ENDF/B-VII.1, the variations of the keff and total power, as the function of the position of the spallation targets, are provided for each multibeam concept. The results show that the keff and total power were affected by an interference effect between the spallation targets. For the transport of fission neutrons in the core, the maximum radius of the interference effect between the spallation targets was 40 cm. Considering the transport of spallation neutrons in the ADS-NWT, the maximum radius of the interference effect between the spallation targets was 60 cm. The spallation targets were moved from the inner circle to the outer circle of the fuel zone, and the different variations in keff and total power trend for the three-beam, four-beam, six-beam, and seven-beam concepts for the ADS-NWT were obtained.

Calculation of the Neutron Parameters for Accelerator-Driven Subcritical Reactors

In this paper, the accelerator-driven subcritical reactor (ADSR) is simulated based on structure of the TRIGA-Mark II reactor. A proton beam is accelerated and interacts on the lead target. Two cases of using lead are considered here: firstly, solid lead is referred to as spallation neutron target and water as the coolant; secondly, molten lead is considered both as a target and as a coolant. The proton beam in the energy range from 115 MeV to 2000 MeV interacts with the lead to create neutrons. The neutron parameters as neutron yield Yn/p, neutron multiplication factor k, the radial and axial distributions of the neutron flux in the core have been calculated by using MCNPX program. The results show that the neutron yield increases as the energies of the proton beam increases. When using the lead target, the differences between the neutron yield are from 4.2% to 14.2% depending on the energies of the proton beam. The proportion of uranium in the mixtures should be around 24% to produce an effective neutron multiplier factor greater than 0.9. The neutron fluxes are much higher than the same calculations for the TRIGA-Mark II reactor model using tungsten target and light water coolant.

Analyzing the Neutron Parameters in the Accelerator Driven Subcritical Reactor Using the Mixture of Molten Pb-Bi as Both Target and Coolant

In this paper, the Accelerator Driven Subcritical Reactor (ADSR) was simulated based on the structure of the TRIGA-Mark II reactor by the MCNPX program. The proton beam interacts on the Pb-Bi molten target with various energy levels from 0.5 GeV to 2.0 GeV. The important neutron parameters to evaluate the operability of ADSR were calculated as: the neutron yields according to various thicknesses of the target and according to the energy of the incident proton beam; the effective neutron multiplication factor for various fuel mixtures, along with its stability for some fuel mixtures; the axial and radial distributions of the neutron flux along with the height and radius of the core. The obtained results had shown a good agreement in using Pb-Bi molten as the interaction target and coolant for ADSR.

Radioactive decay of thorium and uranium in the liquid lead and molten salt

The accelerator driven subcritical reactor (ADSR) is very interested in the recent years. One of the researches is about the fuel, fuel should be thorium or uranium; the coolant is liquid lead or molten salt. This paper presents the results of research on the radioactive spectrum of thorium in molten salt, uranium in molten salt, and liquid lead. These research results include energy spectrums of alpha, beta, gamma, neutrino particles; the energy spectrums of the daughter nucleus.

Dynamic Characteristics of Accelerator-Driven Subcritical Reactor With Self-Adapting Multi-Mode Core Few-Group Constants

In this study, the dynamic characteristics of accelerator-driven subcritical reactor (ADSR) under beam transients with high heterogeneity of neutron flux in time-space are investigated. Multi-mode core few-group constants are generated by three kinds of neutron fluxes: steady state of ADSR, λ-eigenvalue fundamental wave, and α-eigenvalue fundamental wave. The proposed few-group constants overcome the limitation of single few-group constant generated by two-step method that cannot consider the variation in neutron flux density and neutron energy spectrum with time and space. Compared to the existing few-group constants generated by two-step method under different operating conditions, the self-adapting multi-mode core few-group constants exhibit higher accuracy in the case of following two modes: steady-state mode of ADSR in the starting process and λ-eigenvalue mode in the beam trip condition. Overall, this research provides useful insights on neutron kinetics and can boost the development of ADSRs.

Simulating the Production of Medical Radioisotopes in a Fast Thorium-Cycle ADS with SERPENT

Radiopharmaceuticals are used in nuclear medicine for diagnostic or therapeutic acts. The short decay half-lives of medical radioisotopes, especially those used for diagnostics, imply that they should be produced continuously and transported as quickly as possible to the medical units where they are used. Neutron-rich medical radioisotopes are generally produced in research reactors, like technetium-99m, lutetium-177, holmium-166 and iodine-131. On the other hand, proton-rich radioisotopes are produced via reactions with charged particles from accelerators like fluorine-18, gallium-67, iodine-123 and thallium-201. Beside this, innovative nuclear reactors are advocated as solutions to the issues of nuclear waste production and proliferation threats. Fast neutron, thorium-cycle and accelerator-driven subcritical (ADS) reactors are some of the most promising of them, proposed as safer fuel breeders and “waste burners”. This article examines the use of a fast thorium-cycle ADS with liquid lead-bismuth eutectic coolant for the production of molybdenum-99/technetium-99m and lutetium-177. Burnup simulation has been made with the Monte-Carlo (MC) code SERPENT. It is demonstrated that MC codes can advantageously be used to determine the optimal irradiation time for a given radioisotope in a realistic reactor core. It is also shown that fast thorium-cycle ADS is an economical option for the production of medical radioisotopes.

Study of the Effects of Moderators on ADS System Performance Based on UN-ThO2 Fuel

The neutron spectrum is an essential factor in making possible the increase of 233U isotope breeding from thorium fuel in an accelerator-driven subcritical (ADS) system; therefore, studying the effects of various moderators and coolants on 233U breeding is an important step in ADS performance. This study aims to evaluate the effect of using different moderators and coolants on the ADS system characteristics. Sodium, which was the most common coolant used in ADS reactors, was replaced by light water (LW) and graphite + CO2, separately. In this study, we used uranium nitride as the seed fuel associated with ThO2 as the blanket fuel for all cases. The Monte Carlo transport code MCNPX 2.7.0 was used to calculate neutronic parameters such as effective multiplication factor (Keff), power peaking factor (Pmax/Pav) in the radial direction of the ADS reactor core, actinide isotope evolution during fuel burnup, and power fraction from each fuel type for all cases. The results show that the utilization of graphite as the moderator with CO2 as the coolant allows more 233U production in thorium fuel compared with sodium and LW. On the other hand, LW showed great ability for plutonium and minor actinide transmutation and for energy generation.

K effective factor in the ADSR using liquid lead target and (Th233U)O2, (Th235UO)2, (Th238U)O2 fuel mixture

The study of using liquid lead target for the accelerator driven subcritical reactor (ADSR) has been proposed by Nguyen Mong Giao, Vu Thi Diem Hang, Tran Minh Tien (International Journal of Modern Physics and Application. Vol. 2, No. 3, 2015). The calculations about neutron distribution, neutron fluxwith thorium fuel and thorium mixturehas been showed by Tran Minh Tien (Distribution of Neutrons from The Reaction (p, n) on the Liquid Lead Target in The Accelerator Driven System Reactor, J. Phys.: Conf. Ser. 1172 012066, 2019; Distributions of neutron flux from (p, n) reaction on the liquid lead target for accelerator driven subcritical reactor (ADSR), J. Phys.: Conf. Ser. 1324 012061, 2019). In this paper, the k effective factor is calculated with(Th233U)O2, (Th235U)O2and(Th238U)O2fuel mixture.

Development and verification of code IMPC-Depletion for nuclide depletion calculation

Simulation of nuclide inventory evolution in the reactor during the entire irradiation history is of significance to the design of fuel component and core layout scheme and nuclear waste disposal. To perform accurate burnup calculations for fast reactor systems such as accelerator-driven sub-critical reactors, a nuclide depletion code IMPC-Depletion has been developed. Depletion of reactor material is described by thousands of coupled first-order ordinary differential equations, which is the stiff ODEs system. For IMPC-Depletion, algorithms including Chebyshev Rational Approximation Method (CRAM) and Transmutation Trajectory Analysis method (TTA) are implemented to solve Bateman equations. The parameter-extended XML file format is used as standard input. Decay, constant flux and power modes are supported. To verify IMPC-Depletion code, decay of Ra228, fixed-flux irradiation of Zircaloy-4 and 2.4% UO2, fixed-power irradiation of 3.2% UO2 are tested. Results are basically consistent with those of ORIGEN2.1. In addition to independent nuclides inventory analysis, IMPC-Depletion can also be coupled with Monte Carlo neutron transport code to solve burnup problem. Here, simple MOX pin cell and OECD/NEA fast reactor burnup benchmarks are employed to verify the Out-coupling Monte Carlo burnup code system IMPC-Burnup2.1, which couples OpenMC and IMPC-Depletion. Result parameters such as keff, reactivity changes and nuclides inventory values are reliable when compared with other participants.

Calculation of the neutron flux distribution in the accelerator driven subcritical reactor with (Th-233U)O2 and (Th-235U)O2 mix fuel

This paper presents results of calculating the neutron flux distribution in an accelerator driven subcritical reactor (ADSR) with (Th-233U)O2 and (Th-235U)O2mixed fuel. An ADSR consists of 90 fuel rods, and 10 graphite reflector rods. All objects are placed in liquid lead. Thorium is replaced by mixture of (Th-233U)O2 and (Th-235U)O2; MCNP5 program has been used to calculate radial distribution of the neutron flux, axial distribution and energy distribution from (p, n) reaction.The calculation results show that the axial distribution of the thermal and fast neutron flux reduce from the center of core but reduction rate is different.The thermal neutron flux decreases gradually from 0 to 2.5cm; decreases rapidly from 25cm to 5cm. In comparison, the thermal neutron flux is smaller than fast neutron flux from 0 to 4cm along the radius but the thermal neutron flux is larger than fast neutron flux at distances greater than 5cm along the radius of the reactor.

Design of an accelerator-driven subcritical dual fluid reactor for transmutation of actinides

An accelerator-driven subcritical dual fluid reactor (AD-DFR), which is a hybrid core operated by a high power accelerator, is designed for the transmutation of minor actinides. The subcritical core is dual in the sense that a lead-bismuth-eutectic-cooled fast reactor (LFR) is combined with a molten salt reactor (MSR). Thus, the core has two loops: one for the liquid metal coolant and the other for the molten salt fuel. The combination of LFR and MSR can take advantages of both reactor types. A subcritical core allows for loading a high fraction of minor actinides in fuels. An 800MW_t AD-DFR can transmute minor actinides approximately 120kg/year with only the maximum beam power of 13MW.

Distributions of neutron flux from (p, n) reaction on the liquid lead target for accelerator driven subcritical reactor (ADSR)

This paper presents the results of calculating the distributions of neutron flux in an accelerator driven subcritical reactor (ADSR) with thorium fuel. The simulated ADSR consists of 90 fuel rods and 10 graphite reflector rods. All objects are placed in liquid lead. MCNP5 program is used to calculate energy distributions of the neutron flux, axial distributions and radial distributions inside the core of the ADSR. The results show that the neutron fluxes of 250, 450, 600 and 850 angles are similar; axial distributions of neutron flux decrease gradually; the neutron fluxes are the largest between 0.3 and 0.5 MeV. The radial distributions of neutron flux decrease gradually from center, and the largest value is between 30 and 40 cm. The results also show that the neutron fluxes are different with different radii. The results are important for the design and operation of the ADSR.

Verification of a HC-PK-CFD coupled program based a benchmark on beam trip transients for XADS reactor

It is important to employ multi-physics coupling strategy in thermal-hydraulics/safety analysis of nuclear reactors. In general, it is based on couplings of system codes with neutron dynamics codes, which is incompetent to cope with multi-dimensional thermal-hydraulics/safety problems involved in liquid metal cooled pool-type reactors. To cope with it, Computational Fluid Dynamics tools are good choices which have attracted considerable attentions in reactor simulation these years. In order to apply a commercial CFD program in thermal-hydraulics/safety analyses, the research on couplings of Point Kinetics (PK) model and a fuel pin Heat Conduction (HC) model to the commercial CFD platform FLUENT abbreviated as HC-PK-CFD is performed in this study. Verifications are performed by the benchmark simulation on beam interruption transients for an accelerator-driven sub-critical reactor system. It is demonstrated that the results simulated by HC-PK-CFD agrees well with the ones by other ten multi-physics coupling codes participating in this benchmark.

Development and validation of depletion code system IMPC-Burnup for ADS

Depletion calculation is important for studying the transmutation efficiency of minor actinides and long-life fission products in accelerator-driven subcritical reactor system (ADS). Herein the Python language is used to develop a burnup code system called IMPC-Burnup by coupling FLUKA, OpenMC, and ORIGEN2. The program is preliminarily verified by OECD-NEA pin cell and IAEA-ADS benchmarking by comparison with experimental values and calculated results from other studies. Moreover, the physics design scheme of the CIADS subcritical core is utilized to test the feasibility of IMPC-Burnup program in the burnup calculation of ADS system. Reference results are given by the COUPLE3.0 program. The results of IMPC-Burnup show good agreement with those of COUPLE3.0. In addition, since the upper limit of the neutron transport energy for OpenMC is 20 MeV, neutrons with energies greater than 20 MeV in the CIADS subcritical core cannot be transported; thus, an equivalent flux method has been proposed to consider neutrons above 20 MeV in the OpenMC transport calculation. The results are compared to those that do not include neutrons greater than 20 MeV. The conclusion is that the accuracy of the actinide nuclide mass in the burnup calculation is improved when the equivalent flux method is used. Therefore, the IMPC-Burnup code is suitable for burnup analysis of the ADS system.

Proton-induced spallation of mercury and xenon targets for neutron flux produced between CEM03.01 and INCL4-ABLA spallation packages

Accelerator-Driven Subcritical (ADS) reactors are one of the pathways considered for the incineration of minor actinides. They use a proton beam accelerator up to a GeV energy, bombarding a spallation target, usually made of a material of high atomic number (W, Pb, Ta or U). The neutrons generated during the spallation reactions are then multiplied. The main goal of this study is to make a comparison between two models of physical packages via the Monte Carlo code MCNP6. The first package is the INCL4-ABLA and the second one is the CEM03.01, and we have deduced the best package according to the neutron production rate. In this study, we are interested in two cylindrical targets of natural mercury and xenon 20cm in height and 20 in diameter and a proton beam varies from 0.1 to 1.5 GeV.