Subcritical Reactor Research Articles

Embedded fission source approach to analyze external source effect in a subcritical reactor

A simple modification of the fission source spectrum is proposed for the Monte Carlo calculation for externally driven subcritical reactors. Embedded fission source (EFS) approach replaces the fission spectrum with the collective spectrum of the external source and fission working in a small region near to external source only. This model provides eigenvalue search with external source contribution resulting in direct connection with isotope depletion calculation within Monte Carlo simulation. MCNPX 2.6.0 is modified for EFS and tested for three kinds of reactor problems; critical fast reactor driven subcritical reactor, accelerator driven subcritical reactor and fusion-fission hybrid reactor. The effectiveness of the external sources is determined by the extra amount of the TRU isotopes burnt and increase in the core multiplication level. The tests were performed for two subcritical cases; keff ∼ 0.986 and keff ∼ 0.959, with initial TRU loading of 188.8 kg and 183.2 kg, respectively. The EFS approach is validated by simulating a fission source driven system which gave the multiplication factor comparable to the original MCNPX within one standard deviation (32 pcm). The difference in TRU burn-up for both subcritical cases is of the order of 0.1% for each driving neutron source option. The study showed that the EFS can be a simple and a reliable approximation for the external source analysis.

Accelerator driven system for transmutation and energy production

The significant quantities of highly radiotoxic long-lived nuclear waste necessitate the research of methods of their transmutation to short-lived ones. This can be achieved by interaction between nuclear waste and neutrons (n,xn)-reactions and radiative captive). The ADS (Accelerator Driven System) is the most promising tool for such transmutation. ADS consists of powerful high energy accelerator irradiating a target of heavy nuclei: uranium, thorium, lead and bismuth. Inside of the target is generated a high-energy neutron field. Several ADS constructions for waste transmutation and energy production are considered. The influence of the cooling system on the neutron flux and the energy production in ADS are investigated. The radiation and biological protection are also discussed.

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

Optimization Model for Radioactive Waste Transmutation in Advanced Fuel Cycle

Abstract Optimization approach to effective radioactive waste management in advanced fuel cycle is proposed in the paper. This approach is based on the optimal control theory. The considered controlled system contains a system of ordinary differential equations, describing radioactive isotopes concentration change in time and a number of switching points in which the system state can be changed. The values of variables in the switching points are the optimizing parameters, subjected to determination. Moreover, because the recycling of nuclear waste is supposed to fulfill in subcritical reactor, the constraints on effective multiplication factor are also taken into account during numerical calculations.

Investigation of spallation neutron production by 1 GeV protons beam

The MCNP code based on the Monte Carlo model is intended for neutron simulations. It is used to study neutron production, neutron flux calculations, energy deposition, dose, and multiplicity of neutrons; in this paper, we used the MCNP6 code for neutron simulation in spallation targets. We are interested in a 1 GeV proton beam that enters frontal impact with several targets; the geometry of all targets is 10 cm in height and 69 cm in radius. The simulations made in this study clearly show that there are no heavy targets (Co, Cd and Fe), but their neutron production rates are remarkable. We also conclude that the selected targets may be useful to handle almost like spallation targets in an accelerator-driven system.

A new approach on modeling of the b-viii, the ultimate achievement of the second “Uranverain”

German Nazi state conducted researches in nuclear technologies as an attempt to achieve various military goals. As the result of these researches, German scientists developed different, advanced nuclear technologies in years before and during World War II. In an attempt to develop the ?Uranmaschinen?, in which controlled release of high energy in fission process can be achieved, various approaches were examined, theoretically and experimentally. These studies were conducted under support of the German Nazi state and were known as the First and Second ?Uranverain? (Uranium Society/Club). Versions of the ?Uranmaschinen? were based, mainly, on natural uranium fuel and moderators of heavy water, regular water or paraffin. The latest known fission device was the subcritical nuclear fission reactor B-VIII, re-built in village Haigerloch, Bavaria, Southern Germany, in first months of 1945. It was a tank type device with natural uranium metal fuel and heavy water moderator, reflected by graphite. Radiation shielding of the device was achieved, primarily, by surrounding the reactor tank by regular water. The whole device construction was assembled inside a concrete hole in the floor of an underground cave, ex beer cellar. A recent neutronics study of this reactor was done, assuming fuel rods with lumped parameters approximation, by Italian Bologna University LIN (Laboratorio Ingegneria Nucleare) research group in 2009. This paper is a new approach to the neutronics study of the B-VIII reactor with an attempt to model real fuel-moderator geometry. This study points out many approximations and simplifications, made during the B-VIII material composition and geometry modeling, due to missing data. The paper investigates the influence to criticality of numerous uncertainties in the material compositions, mass densities and geometry of the facility. The Monte Carlo MCNP6.1 code with the latest ACE type neutron nuclear cross section data is used for that purpose. Additionally, an attempt of estimation of the uncertainty of the experimental result of the neutron multiplication was given. Differences in the calculated values of the neutron multiplication and the experimental one are investigated and tried to explain. These analyses show that the B-VIII was a subcritical device, as it was shown by the experimental results of the German scientists achieved in March-April 1945 in Haigerloch. <br><br><font color="red"><b> This article has been corrected. Link to the correction <u><a href="http://dx.doi.org/10.2298/NTRP1802230E">10.2298/NTRP1802230E</a><u></b></font>

Determination of Design Extension Conditions List and Design Optimization of Safety Systems for ADS by Lines of Defenses Method

During the conceptual design phase of an accelerator-driven sub-critical system (ADS), the reliability requirements of safety systems need to be determined for future practical construction to balance the economy performance with the safety performance. In this paper, the design optimization of the safety-related systems reliability for the ADS based on China LEAd-based Research Reactor was performed by applying the lines of defense (LODs) method to this ADS. First, a tentative LOD type, such as a strong line (a LOD) or a medium line (b LOD), was assigned to the main safety related systems. These systems include the accelerator trip system, decay heat removal system, reactor vessel, secondary cooling system isolation, decompression system, emergency diesel engine, and containment systems. Second, the failure of the LODs were combined with initiating events that were identified early, and a preliminary list of design extension conditions (DECs) was obtained. Third, the suitability of the conditions for classification into the design basic condition (DBC) or the DEC and their unacceptable consequences for the DBC or criteria limitation were discussed. Last, the LOD types of safety systems were re-assigned considering the results to optimize the design. The complementary events in the list of DECs are obtained. This paper can provide important references for the construction and operation of future ADSs.

Design of fast adaptive readout system for wire scanners

A new wide-range fast readout system capable of adaptive identification is designed for wire scanners, which are used to measure beam profiles and emittance. This system is capable of handling varying current signals with Gaussian distributions and current pulses up to 1000 counts/s, as well as an input current range of 1 nA–1 mA. When tested, the resolution was found to exceed 3.68% for full scale, the nonlinearity was found to be less than 0.11%, and the measurement sensibility was found to be less than 5 pA. We believe that the system will play a crucial role in improving the measurement accuracy of beam diagnosis and the efficiency of accelerator operation, as well as decreasing the time required for beam tuning. This system was applied to the beam diagnosis of an injector II prototype for an accelerator-driven subcritical system and produced excellent measurement results. A description of the adaptive fast readout system for wire scanners is presented in this paper.

CAD modeling study on FLUKA and OpenMC for accelerator driven system simulation

Accelerator driven system (ADS) is an advanced nuclear system with high inherent safety feature, which can transmute the nuclear waste including minor actinides (MA) and long-lived fission products (LLFPs) into short-lived fission products. In order to accomplish the neutronic simulation in ADS, a conventional way can be divided into two steps, namely the physical process of proton beam bombarding spallation target (FLUKA) and the neutron transport process in subcritical reactor coupling with a spallation neutron source (OpenMC). However, there are some difficulties in the modeling process for ADS, such as the inconsistent modeling process between FLUKA and OpenMC codes, and the error-prone process of constructing complex geometries. Therefore, it is imperative to build a code system that can convert the CAD modeling files into the input cards of Monte Carlo codes. In this context, a code system named CAD-PSFO (FreeCAD based parsing script for FLUKA and OpenMC) has been developed to solve the modeling conversion problem in ADS, in which basic geometry and boolean logic classes have been established with a mapping relationship to FLUKA and OpenMC codes, and ray-casting technology based iteration method has been proposed to solve the problem of higher-order surfaces in complex geometries. Finally, CAD-PSFO has been verified and validated by comparing with a reference model, two benchmarks, and two simulation tests.

МОДЕЛЮВАННЯ РОЗПОДІЛУ ПАЛИВА У ВНУТРІШНІЙ ШВИДКІЙ ЗОНІ ДВОЗОННОГО ПІДКРИТИЧНОГО ЯДЕРНОГО РЕАКТОРА

МОДЕЛЮВАННЯ РОЗПОДІЛУ ПАЛИВА У ВНУТРІШНІЙ ШВИДКІЙ ЗОНІ ДВОЗОННОГО ПІДКРИТИЧНОГО ЯДЕРНОГО РЕАКТОРА

Intensity limits of the PSI Injector II cyclotron

We investigate limits on the current of the PSI Injector II high intensity separate-sector isochronous cyclotron, in its present configuration and after a proposed upgrade. Accelerator Driven Subcritical Reactors, neutron and neutrino experiments, and medical isotope production all benefit from increases in current, even at the ∼10% level: the PSI cyclotrons provide relevant experience. As space charge dominates at low beam energy, the injector is critical. Understanding space charge effects and halo formation through detailed numerical modelling gives clues on how to maximise the extracted current. Simulation of a space-charge dominated low energy high intensity (9.5 mA DC) machine, with a complex collimator set up in the central region shaping the bunch, is not trivial. We use the OPAL code, a tool for charged-particle optics calculations in large accelerator structures and beam lines, including 3D space charge. We have a precise model of the present (production) Injector II, operating at 2.2 mA current. A simple model of the proposed future (upgraded) configuration of the cyclotron is also investigated. We estimate intensity limits based on the developed models, supported by fitted scaling laws and measurements. We have been able to perform more detailed analysis of the bunch parameters and halo development than any previous study. Optimisation techniques enable better matching of the simulation set-up with Injector II parameters and measurements. We show that in the production configuration the beam current scales to the power of three with the beam size. However, at higher intensities, 4th power scaling is a better fit, setting the limit of approximately 3 mA. Currents of over 5 mA, higher than have been achieved to date, can be produced if the collimation scheme is adjusted.

Beta-effective sensitivity and uncertainty analysis of MYRRHA reactor for possible use in nuclear data validation and improvement

New reactor concept studies, such as those of the MYRRHA Accelerator Driven System (ADS), with their extended nuclear data needs relaunched the interest in sensitivity and uncertainty (S/U) studies. S/U analysis play an essential role in assuring the accuracy and reliability of nuclear reactor computations and provide an insight in the physical phenomena involved in the neutron transport. This paper presents the study of the MYRRHA critical core, which is designed to operate at 100 MW thermal power. The previously presented analysis of the MYRRHA effective multiplication factor (keff) was extended to study the sensitivity and uncertainty in kinetic parameters, such as the effective delayed neutron fraction (βeff). The sensitivity and uncertainty computations were performed by means of the SUSD3D code based on forward and adjoint-flux first-order perturbation theory. The sensitivity coefficients of βeff with respect to the basic nuclear data were calculated by the k-ratio derivation method proposed in 2011. The main components of the uncertainties in βeff were estimated using the JENDL-4.0u covariance data. The energy dependent sensitivity profiles of keff and βeff with respect to nuclear data are compared in view of exploiting the differences among them for a physically more complete, comprehensive and consistent nuclear data validation and improvement procedure. The accuracy of the keff and βeff measurements and calculations has to be carefully evaluated in order to conclude on the nuclear data status and possible improvements.

Reactivity determination using the hybrid transport point kinetics and the area method

This paper presents an application of the hybrid transport point kinetic (HTPK) technique to the reactivity determination in subcritical reactor configurations. The mathematical model of the HTPK, initially proposed by Picca et al. (2011) to simulate the time-dependent neutron transport, is here extended to incorporate delayed emissions. The classical area method (Sjӧstrand, 1956), developed to invert the point kinetic (PK) model, is then adapted to accommodate the peculiarities of the HTPK approach, to allow its analytical inversion. This novel inverse neutron kinetic methodology is tested on a three-region reactor configuration, showing the interesting performance of the approach based on the HTPK model as compared to the standard area method and highlighting its potential to overcome some of the limitations of the PK-based inversion.

Experimental analysis and uncertainty quantification using random sampling technique for ADS experiments at KUCA

ABSTRACT Nuclear data-induced uncertainties of neutronics parameters (neutron multiplication factor keff, one-point kinetics parameters and prompt neutron decay constant α) are quantified for lead-bismuth zoned accelerator-driven system experiments at the Kyoto University Critical Assembly, in order to contribute validation for subcritical core analysis. The random sampling technique using SCALE6.2.1/Sampler/NEWT/PARTISN is utilized for the validation and the uncertainty quantification, because the random sampling technique is applicable for a problem which is not easy to apply the perturbation theory. Consequently, it is confirmed that the numerical results of α reasonably agree with the experimental ones, compared with the nuclear data-induced uncertainties. In addition, it is clarified that the nuclear data-induced correlations between α and keff and between α and neutron generation time Λ are strongly negative and positive, respectively. This fact implies that the numerical predictions of keff and Λ can be improved by the data assimilation technique using subcritical experimental results of α, which can be directly measured even for a deep subcritical system.

Neutronic design study of accelerator driven system (ADS) for Jordan subcritical reactor as a neutron source for nuclear research

In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor Keff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux.

Influence of gamma-ray irradiation on mechanical property of YSZ for oxygen sensors in ADS

In the accelerator driven system, oxygen sensors, on which yttria-stabilized-zirconia (YSZ) is used for the electrolyte, will be irradiated by gamma-ray in activated lead bismuth eutectic (LBE) adopted as a both coolant and spallation target. Partially stabilized zirconia, like a 3Y-YSZ, has high strength due to the stress induced phase transformation from tetragonal (T) to monoclinic (M) phase. The mechanical properties of YSZ specimens after gamma-ray irradiation, thermal aging and immersion in LBE have been investigated in terms of the fracture surface structure. In the case of 6Y and 8Y-YSZ, the 4-point bend strength did not change even after gamma-ray irradiation and thermal aging. The 3Y-YSZ, however, showed that the strength was decreased a little by the gamma-ray irradiation, thermal aging and also immersion in LBE. The XRD results indicate that in the case of 3Y-YSZ after the annealing and soaking in LBE, the pristine surface, which is not fractured and strained by bending test, shows M phase much more than the surface of as-received one. After surface polishing of 3Y-YSZ the XRD intensity of M phase increased with increasing gamma-ray dose for the pristine surface. The bend strength of irradiated 3Y-YSZ after polishing, however, did not change in this study. These results suggest that the process of gamma-ray irradiation and thermal process around 450 °C cause pre-phase transformation from T to M, which prevents the stress induced phase transformation previously.

Reaction rate analyses of accelerator-driven system experiments with 100 MeV protons at Kyoto University Critical Assembly

ABSTRACT At the Kyoto University Critical Assembly, a series of reaction rate experiments is conducted on the accelerator-driven system (ADS) with spallation neutrons generated by the combined use of 100 MeV protons and a lead–bismuth target in the subcritical state. The reaction rates are measured by the foil activation method to obtain neutron spectrum information on ADS. Numerical calculations are performed with MCNP6.1 and JENDL/HE-2007 for high-energy protons and spallation process, JENDL-4.0 for transport and JENDL/D-99 for reaction rates. That the reaction rates depend on subcriticality is revealed by the accuracy of the C/E (calculation/experiment) values. Nonetheless, the accuracy of the reaction rates at high-energy thresholds remains an important issue in the fixed-source calculations. From reaction rate analyses, the indium ratio is newly defined as another spectrum index with the combined use of 115In(n, γ)116mIn and 115In(n, n′)115mIn reaction rates, and considered useful in examining the neutron spectrum information on ADS with spallation neutrons.

Kinetics solution with iteration scheme of history-based neutron source in accelerator driven system

In the time-dependent neutronic analyses of the accelerator driven system (ADS), one of the technical issues is the investigation method of reactor transients for validating why detector responses or fission rates are varied in functions of space and time. In previous studies, numerical analyses for ADS have been conducted by using forward transport methods, focusing on reconstruction of the time behavior. In this study, a source history-based kinetics equation, which is superior to cause analysis of time transients comparing with general forward transport methods, is derived by using the time contributions from each source into responses. For the validation of the applicability, through the re-evaluation of ADS experiment at the Kyoto University Critical Assembly, the time transients of fission rates and detector responses are quantitatively analyzed with the time contributions directly obtained by the proposed method. The expanded analysis ability provided by the proposed method will contribute to investigate the reactor transient of ADS including verification of reactor experiments, analysis of space-time detector responses and establishment of estimation strategy for obtaining the kinetics parameters.

Effect of high-energy neutron source on predicting the proton beam current in the ADS design

Abstract The accelerator-driven subcritical system (ADS) is driven by a neutron source from spallation reactions introduced by the injected proton beam. Part of the neutron source has energy as high as a few hundred MeV to a few GeV. The effects of high-energy source neutrons (E n > 20 MeV) are usually approximated by energy cut-off treatment in practical core calculations, which can overestimate the predicted proton beam current in the ADS design. This article intends to quantize this effect and propose a way to solve this problem. To evaluate the effects of high-energy neutrons in the subcritical core, two models are established aiming to cover the features of current experimental facilities and industrial-scale ADS in the future. The results show that high-energy neutrons with E n > 20 MeV are of small fraction (2.6%) in the neutron source, but their contribution to the source efficiency is about 23% for the large scale ADS. Based on this, a neutron source efficiency correction factor is proposed. Tests show that the new correction method works well in the ADS calculation. This method can effectively improve the accuracy of the prediction of the proton beam current.

Preliminary sustainable and economic analysis of nuclear fuel cycle with subcritical system

Summary Nuclear power is attractive for helping carbon reduction and energy restructuring, but the low utilization of nuclear resource and technical problems on reprocessing are key issues for further nuclear development. Subcritical system might be a potential way to solve these problems, which include fusion-driven subcritical system and accelerator-driven subcritical system. Especially the new subcritical system conception spent fuel burner could be used for waste transmutation, nuclear fuel breeding, and energy production. In this contribution, 2 nuclear fuel cycle scenarios with subcritical system for spent fuel burning and pressurized water reactor were assessed with the performances of sustainability of resources, environmental impacts, and economy. The results showed that the advanced fuel cycle based on subcritical system is an attractive option for the long-term development of nuclear power in China.