Burnup Reactivity Research Articles

Fuel cycle features of RFBB fast reactor with advanced fuel

ABSTRACT The advancement of innovative fast nuclear reactor systems offers a dependable and sustainable solution to address the world’s growing energy needs. Breed-and-burn systems, in particular, have the potential to efficiently utilize uranium fuel, which could lead to a reduction in nuclear waste and eliminate the need for expensive fuel reprocessing facilities. In this paper, we design a core for a Rotational Fuel-shuffling Breed-and-Burn reactor with Silicide fuel and Sodium coolant (RFBB-SS). To demonstrate the core’s fuel cycle features, we then analyze the reactor burnup parameters through neutronic calculations. Then, to reveal its safety features, the control rod assembly worth and the maximum temperatures of the reactor domains were estimated. The results showed that a small start-up core for an RFBB-SS design is feasible and that the maximum temperatures of the fuel cell domains in a hot channel can be kept well below their safety limits. Hence, the reactor could achieve a subcritical state when all control rod assemblies are inserted into the core.

Increased 238Pu Production in Proliferation-Resistant U-10Zr Fuel Produced from Reprocessed Uranium Containing Unseparated 237Np

The sodium-cooled fast reactor (SFR) is a potential candidate for the future production of clean sustainable energy. Some SFR designs planned for construction will use high-assay low-enriched uranium (HALEU) metal fuel enriched up to around 19.75%. However, fast reactor technology faces several criticisms, such as its ability to produce weapons-grade plutonium, the possibility of facility diversion from peaceful use to the production of nuclear weapons–usable materials, and the risk of incentivizing the construction of domestic fuel enrichment programs. To compete with traditional light water reactors (LWRs), SFR technology utilizing HALEU fuel should be developed to be proliferation-resistant and sustainable. In this study, the Serpent 2 Monte Carlo reactor physics code was used to simulate and compare alternative HALEU fuel options that could be implemented to reduce proliferation risk. As part of the comparison, nonpeaceful plutonium diversion pathways were considered. By evaluating its attractiveness for diversion from peaceful use to the production of nuclear weapons–usable material, it was shown that the proliferation resistance of SFR technology utilizing HALEU fuel can be improved. The operation of a once-through fuel cycle SFR utilizing HALEU fuel was simulated, and the performance and material attractiveness of various U-10Zr fuel options were compared. Currently, hundreds of thousands of tons of spent fuel from the operation of LWRs are stored across the world. Most of the heavy metal in this spent nuclear fuel is comprised of reusable uranium and plutonium. Re-enrichment of reprocessed uranium (RepU), with or without the presence of unseparated 237Np, for the production of SFR fuel is one way to simultaneously improve uranium resource utilization and enhance proliferation resistance due to the co-enrichment of 236U and 237Np. The presence of these isotopes in fresh HALEU fuel can increase 238Pu production even at low reactor burnup. Due to its high decay heat and spontaneous neutron emission, 238Pu decreases material attractiveness for nuclear weapons construction. Directly doping fresh fuel with preseparated 237Np requires the separate storage of neptunium, introducing its own nuclear security requirements and proliferation risks. Additionally, due to uncertainty in the separation factor between RepU and neptunium, some neptunium impurities will inevitably remain in RepU following separation. Therefore, this study examined the impact of unseparated 237Np on the production and depletion of U-10Zr fuel. A parametric study comparing fuel options of varying 237Np weight fractions was conducted, and proliferation scenarios considering both clandestine and overt diversion of discharged fuel were considered.

Minor actinides transmutation in pressurized water reactors. 2. Using uranium and thorium fuel to burn minor actinides in a system with several VVER reactors

There is currently a consensus among the scientific and engineering community regarding the solution to the problem of minor actinides (MAs) formed in the process of nuclear power operation: MAs need to be converted to fission products during burnup in power reactors. Fast neutron reactors (BN, BREST) and molten salt reactors (MSR) are considered largely to this end. Despite the advantages of using fast reactors, there are no currently power unit designs with BN or BREST reactors available for commercial operation. The possibility of using VVER reactors for this purpose is rarely covered in scientific literature, despite the fact that the technology of light water reactors has long been mastered, and preparing a VVER-based MA burner reactor is technically simpler than on the basis of a pilot commercial BN technology or BREST or MSR reactors at different R&D stages. In the previously published first part of the study by Kazansky and Karpovich 2022, the fuel cycle for VVER reactors was investigated: minor actinides obtained during reactor operation were extracted from spent nuclear fuel and added to fresh fuel for the same reactor. It has been found that implementing this cycle and bringing the concentration of MAs up to 4 wt. % makes it possible to reduce the amount of minor actinides produced in the VVER reactor by a factor of 8, and without a loss in the NPP unit power generation. This paper investigates, based on an idea of closing the fuel cycle in terms of minor actinides, the relationship between the MA burnup depth in a VVER-1200 reactor and the enrichment of fresh fuel, the dynamics of unloaded heavy nuclei, and the amount of MAs in additionally loaded fuel. Under investigation are two types of additionally loaded fuel with oxides of minor actinides added to it: based on enriched uranium dioxide or a mixture of (1 – x) 232ThO2+x (96% UO2) (Kazansky et al. 2023). At the same time, the number of heavy nuclei in fuel does not change when there is a change in the number of loaded MA nuclei, which is measured in the number of the VVER-1200 reactors “served” (the mass of accumulated MA nuclei is about 1% and is removed annually from spent fuel). The number of minor actinides entering fuel is kept at a level that allows having a reactivity margin for the reactor operation at a power of over 2% before the annual refueling. Calculations were undertaken using a fuel assembly model with fuel elements broken down into a number of groups with different fuel burnup depths to simulate the actual loading of the reactor core. This model makes it possible to avoid the power peaking effects and give an answer about the neutronic characteristics of the fuel cycle involving MAs. The calculations made it possible to make a number of important conclusions: more refueling cycles lead to a dynamic equilibrium taking place between the minor actinide amounts loaded and burnt; the number of the reactors served is directly proportional to the enrichment of the make-up fuel and a 20% UO2 enrichment makes it possible to serve up to 23 VVER-1200 reactors, while using 80% 232ThO2 + 20% (96% UO2) fuel allows 17 VVER-1200 reactors to be served; using MAs extracted from SNF with a burnup of 20 to 50 MW*day/kg leads to the MA composition effect on the reactivity balance being within 4 to 5 βeff.

JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient ΔρBU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

Multi-batch core design study for innovative small modular reactor based on centrally-shielded burnable absorber

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.

Bayesian uncertainty quantification of tristructural isotropic particle fuel silver release: Decomposing model inadequacy plus experimental noise and parametric uncertainties

Tristructural isotropic (TRISO) particle fuel is one of the most promising fuel concepts enabling high temperature and high burnup reactor operation. One dominant source of radioactivity released from the TRISO particles is silver (Ag), which is subject to a high release fraction and long decay life compared to other fission products. Previous modeling efforts using the fuel performance code BISON indicated nonnegligible uncertainties in modeling the diffusion process of fission products in TRISO compared to the Advanced Gas Reactor experiments. The overall uncertainties observed when modeling the fission product diffusion can result from uncertainties in model parameters, noisy experimental measurements, and deficiencies in the developed models. The three types of underlying uncertainties have not yet been properly quantified in open literature. This paper presents the Bayesian uncertainty quantification (UQ) using massively parallelizable Markov chain Monte Carlo samplers. The uncertainties due to model parameters, model inadequacy, and experimental measurement noise are quantified, with the σ term used to represent the sum of the model inadequacy and measurement noise uncertainties. It is worth noting that this is the first time the σ term is inferred for nuclear fuel experiments, as compared to using prescribed values for uncertainty quantification in previous work. The parallelizable Markov chain Monte Carlo samplers efficiently infer the model parameters and the σ term, giving insight into physical parameters like diffusion coefficients and the combined model discrepancy and measurement noise. A subsequent forward uncertainty quantification (UQ) is also performed based on the calibration results to generate more accurate predictions of the Ag release. The model inadequacy plus experimental noise is the most dominant source of uncertainty compared to the parametric uncertainty. All the UQ analyses presented in this work are based on the second series of the irradiation experiments in the Advanced Gas Reactor program.

Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility–Part 3: Thermal hydraulic evaluation of large break LOCA under high-burnup conditions

Increasing the peak rod average burnup of pressurized water reactor (PWR) fuel beyond 62 GWd/tU may increase fuel fragmentation, relocation, and dispersal (FFRD) susceptibility during a large break loss of coolant accident (LBLOCA). TRACE thermal hydraulic (TH) LBLOCA analyses were performed for a realistic 24-month high-burnup PWR equilibrium cycle, to inform subsequent transient BISON high-burnup FFRD susceptibility evaluations. Realistic LBLOCA systems behavior was first established by configuring to and comparing with the BEMUSE OECD LBLOCA benchmark. Fuel and operating conditions were then applied from high-burnup VERA depletion calculations. LBLOCA simulations were performed for 281 selected high-burnup rods, for which transient TH boundary conditions were collected for later use in BISON. The TRACE results indicated that rod linear heat rate (rather than burnup) is the main predictor of peak cladding temperature (PCT) during the event. PCT typically occurred at a local burnup lower than the rod-average burnup, especially for twice-burned fuel.

Effect of additive (Zr) and dopants (Pd & Sb) on the depth of migration of lighter lanthanides (Nd & Ce) from U-bearing metal alloys into T91 steel

The metal alloy fuel, which is considered as the advanced nuclear fuel, has its own limitations; namely, lower liquidus temperature, higher swelling, and is highly prone to breach the cladding due to fuel-cladding chemical interaction (FCCI). The mobile and insoluble fission products (FPs), especially lanthanide FPs are mainly responsible for FCCI. In the present study, the influence of additive (Zr) and dopants (Pd and Sb) on the depth of migration of two lanthanide FPs, namely Nd and Ce, from different uranium-bearing metal alloys into T91 (ferritic-martensitic steel) cladding material is investigated. Seven metal alloys, namely, U-Nd, U-Nd-Pd, U-Ce, U-Ce-Pd, U-Zr-Nd, U-Zr-Nd-Pd, and U-Zr-Nd-Sb, were prepared using arc melting technique. The amount of lanthanide FPs taken in the above alloys corresponds to about 10 at.% burn-up of a typical fast breeder reactor (FBR) fuel. The pulsed laser deposition (PLD) method was used to obtain alloys in the form of thin film on T91 cladding, with simultaneous in-situ annealing. Subsequently, these films were subjected to elemental depth profiling using secondary ion mass spectrometry (SIMS). The depth of migration for Nd and Ce from the seven metal alloy systems into T91 cladding was determined based on SIMS data. The extent of the individual and simultaneous effect of additive and dopant on the depth of migration was determined. Both dopants (Pd and Sb) are effective individually in binding Nd and Ce within the fuel matrix (in the U matrix) by forming stable inter-metallic compounds. The effect of Zr on the migration of lanthanide FPs is negligible in the presence of dopants. Stable inter-metallic phases responsible for binding lanthanides FPs are identified. A comparative study on the efficiency of two dopants (Pd and Sb) is also discussed. For the first time, the depth of migration of Nd and Ce from U metal alloys into T91 cladding was measured in the presence of dopants (Sb/Pd) by using a novel technique of PLD followed by SIMS. These studies are very useful for understanding FCCI.

Experimental study on fuel depressurization and its effect on dispersion

Most operating power reactors use oxide fuel in the form of uranium oxide (UO2) which is robust against high temperatures but suffers from poor mechanical performance at high levels of burnup. Currently, the maximum average burnup in light water reactors in the US is limited to prevent severe fuel damage during a hypothetical accident, most importantly the loss-of-coolant accident (LOCA). During a LOCA, high-burnup fuel has been observed to undergo extensive fragmentation, relocation, and dispersal through ruptured cladding, which has the potential to impact both the coolable geometry and long-term cooling of a reactor during such an accident. This paper presents the results from a series of experiments and denotes the approach used to collect such data in which test rods in a mock fuel assembly are filled with a surrogate nuclear fuel and subject to large pressure transients. The resulting deposition of fuel in adjacent subchannels are measured. The results of this study provide an improved method for estimating fuel mass distribution in a reactor core during a LOCA.

A safety enhanced sodium-cooled MOX fueled fast reactor core concept

ABSTRACT We have developed a 750MWe sodium-cooled MOX fueled fast reactor core with enhanced safety against typical ATWS. Design goals were set as follows: 1) negative void reactivity for the loss of flow (ULOF); and 2) burnup reactivity less than 1 $ for the transient over power (UTOP). To achieve the goals the following were considered: 1) adoption of the axially heterogeneous core with a sodium plenum and a gas expansion module to reduce void reactivity; and 2) addition of minor actinides (MAs) to the internal blanket to reduce burnup reactivity. Configurations of the core were optimized as follows. First, void reactivity was reduced by a parameter survey for the inner core height, B-10 content of the upper shield, and standby position of the backup control rods. Second, burnup reactivity was reduced by a parameter survey for the MA content in the internal blanket, etc. We designed the safety enhanced fast reactor core that did not undergo sodium boiling if the ULOF occurred and we maintained fuel integrity for the UTOP by the transient analysis. By installing a self-actuated shutdown system in the backup control rods, the improved operational reliability of this shutdown system and the improved safety can be expected.

Uncertainty Propagation of Fission Product Yields from Uranium and Plutonium in Pebble-Bed HTGR Burnup Calculation

Quantifying fission product yield uncertainty contribution to reactor burnup calculation is an important aspect of pebble-bed High Temperature Gas-cooled Reactor (pebble-bed HTGR) uncertainty analysis. In this work, uncertainty propagation of fission product yield to pebble-bed HTGR burnup calculation is conducted. Uncertainty of fission product yields from four fissile isotopes, namely 233U, 235U, 239Pu and 241Pu, are considered. The stochastic sampling-based uncertainty analysis method is adopted and fission product yield covariance matrices are estimated from ENDF/B-VII.1. The covariance matrix for each fissile actinide is estimated based on the Bayesian method and fission product yields are assigned with log-normal distribution in the sampling process with the Latin Hypercube Sampling (LHS) method. Since the fission fraction from 239Pu plays an important role in fissions of fuels with high burnup value in pebble-bed HTGR, the fission product yield uncertainty contribution from 239Pu is highlighted in this work. The result shows that, in the burnup equilibrium state of pebble-bed HTGR, fission product yield uncertainty contributions from 235U and 239Pu to relative uncertainty of keff are 0.027% and 0.026%, respectively. The overall uncertainty contribution from four fissile isotopes (233U, 235U, 239Pu and 241Pu) to relative uncertainty of equilibrium core keff is 0.038%. Furthermore, fission product yield uncertainty has an important contribution to the nuclide density uncertainty of fission products. The most relative uncertainty, 10.82%, is observed in 109Ag contributed from the fission product yield uncertainty of 239Pu at the burnup equilibrium state. This indicates the uncertainty contribution from the fission product yield of 239Pu cannot be neglected in pebble-bed HTGR burnup uncertainty analysis.

Neutronics design optimization of a small modular fast reactor based on response surface methodology

RSM (response surface methodology) and multi-objective optimization are employed to optimize the preliminary design of a Small Modular Fast Reactor. The basic idea of RSM is to use mathematical models to represent and describe the complicated relationship between inputs and outputs for multi-variable and multi-objective problems. In this study, by means of RSM, a preliminary core design that meets the requirements of both small excess reactivity and small burnup reactivity swing is proposed without empirical preliminary calculation and time-consuming search. The core design is then fine-tuned according to the common engineering requirements. Finally, a 50MWt core design with a burnup reactivity swing<1$ in 20 years has been proposed. In addition, two independent control systems are designed and performance of the core has been studied. Results of the basic core design using RSM show that RSM is a candidate method feasible for preliminary core design.

Mitigating the numerical xenon instability in deterministic reactor burnup calculations

Depletion methods used in modern core simulators for thermal-spectrum reactor burnup calculations may be predisposed to activating, under the correct conditions, a numerical instability that presents itself in the form of artificial spatial xenon oscillations. In this paper a three-dimensional pressurised water reactor depletion problem is utilised to demonstrate that one of the most widely used predictor-corrected depletion schemes can be afflicted by numerical xenon instabilities in reactor burnup calculations performed with a state-of-the-art nodal diffusion core simulator. A method which results in tightening the algorithmic coupling between the 135Xe transmutation and the neutron flux computation steps is proposed for mitigating the numerical xenon instability. The method is called the Xenon Transmutation Feedback (XeTF) method because it incorporates 135Xe transmutation in the feedback iterations of a neutronics solver. The effectiveness of the method is verified by numerical evaluation of some typical reactor depletion scenarios.

Conceptual core design of the molten metal fueled microreactor with self-regulating capability

The microreactor concept has received significant attention in the United States for its lower capital investment, siting flexibility and high mobility. A microreactor aims to provide reliable electricity in remote communities, mining areas, or isolated islands, which will save the high fuel transportation cost. Thus, minimizing the staff level for operating and maintaining a microreactor is desirable. The objective of the study is to design a self-regulated Molten Metal fueled microReactor core (MMR) and to demonstrate the self-regulating capability of the MMR core. The control of the reactor is solely relying on reactivity feedbacks from fuel temperature perturbation. Molten UMn (liquid form at the temperature over 720 °C) was utilized as the fuel material in MMR due to the larger thermal expansion coefficient compared to the conventional solid fuel. The MMR core has a fast neutron spectrum, which corresponds to a small burnup reactivity swing. The core design parameters were selected based on DAKOTA, REBUS, and ANLHTP calculations. The MMR is designed to run for 10 years at a power level of 15 MWth. No refueling is needed during the reactor lifetime. The system dynamic analyses were performed, and it was found that the reactor can be self-regulating within a temperature range of 800 ± 80 °C for a loss of heat sink transient.

Simulating the fuel cycle of a lead-cooled fast reactor

The development of nuclear power with fast reactors is associated with the implementation of a closed nuclear fuel cycle (CNFC). In this regard, one actual task is to simulate the stages of the fuel cycle with study of the neutron-physical characteristics of the core. The design of a reactor for operation in the closed nuclear fuel cycle mode is impossible without the using of verified and certified software packages for calculating fast reactors, capable of simulating all stages of the operation of the reactor facility and the fuel cycle. For the calculations, the FACT-BR software package was used, which has all the necessary capabilities to simulate the operation of the reactor in the closed nuclear fuel cycle mode, taking into account the stages of fuel storage and refabrication. The article presents a technique for modeling the fuel cycle, implemented for the operation of fast reactors with a lead coolant. To demonstrate methodology, a closed nuclear fuel cycle was simulated for the BREST-OD-300 and BR-1200 reactors for the design life. The article describes the scenarios in which the calculation of the burnup of reactor was carried out. In the considered scenarios, it is assumed that the unloading of fuel at the end of the micro campaign is conducted according to the maximum burnup. During the computational modeling the ranges of changes in fuel density and enrichment, reactivity margin, breeding ratio and isotopic composition of plutonium were determined.

Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly computational benchmark using OpenMC Code

A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency. To study &amp;amp; scrutinize the characteristics of one of the VVER-1000 LEU &amp;amp; MOX assembly benchmarks in different states were considered. In this work, the VVER-1000 LEU and MOX Assembly computational-benchmark exercises are performed using the OpenMC software. The work was intended to test the preciseness of the OpenMC Monte Carlo code using nuclear data library ENDF/B-VII.1, against a handful of previously obtained solutions with other computer codes. The kinf value obtained was compared with the SERPENT and MCNP result, which presented a very good similarity with very few deviations. The kinf variation with respect to burnup upto 40 MWd/kgHM was obtained for State-5 by using OpenMC code for both the LEU and MOX fuel assembly. The depletion curves of isotope concentrations against burnup upto 40 MWd/kg/HM were also generated for both the LEU and MOX fuel assembly. The OpenMC results are comparable with those of benchmark mean values. The neutron energy vs flux spectrum was also generated by using OpenMC code. Based on the OpenMC results such as kinf, burnup, isotope concentrations and neutron energy spectrum, it is concluded that the OPenMC code with ENDF/B-VII.1 nuclear data library was successfully implemented. It is planned to use OpenMC code for calculation of neutronics and burnup of the VVER-1200 reactor to be commissioned in Bangladesh by 2023/2024.

The CANDLE burnup strategy applied to small modular pressurized water reactor loading with fully ceramic microencapsulated fuel

Abstract For post-Fukushima nuclear power plants, there has been interested in accident-tolerant fuel (ATF) since it has better tolerant in the event of a severe accident. The fully ceramic microencapsulated (FCM) fuel is one kind of the ATF materials. In this study, the small modular pressurized water reactor (PWR) loading with FCM fuels was investigated, and the modified Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor (CANDLE) burnup strategy was successfully applied to such compact reactor core. To obtain ideal CANDLE shape, it’s necessary to set the infinity or enough length of the core height, but that is impossible for small compact core setting infinity or enough length of the core height. Due to the compact and finite core, the equilibrium state can only be maintained short periods and is not obvious, other than infinitely long active core to reach the long equilibrium state for ideal CANDLE. Consequently, the modified CANDLE shape would be presented. The approximate characteristics of CANDLE burnup are observed in the finite and compact core, and the power density and fuel burnup are selected as main characteristic of modified CANDLE burnup. In this study, firstly, lots of optimization schemes were discussed, and one of optimization schemes was chosen at last to demonstrate the modified CANDLE burnup strategy. Secondly, for chosen compact small rector core, the modified CANDLE burnup strategy is applied and presented. Consequently, the new characteristics of this reactor core can be discovered both in ignition region and in fertile region. The results show that application of CANDLE burnup strategy to small modular PWR loading with FCM fuels suppresses the excess reactivity effectively and reduces the risk of small PWR reactivity-induced accidents during the whole core life, which makes the reactor control more safety and simple.

Analysis of Fuel Burnup and Transmutations at High Burnup of Sodium Fast Breeder Reactor

In this paper, the Monte Carlo N-Particle extended computer code (MCNP) were used to design a model of the European Sodium-cooled Fast Reactor. The multiplication factor, conversion factor, delayed neutrons fraction, doppler constant, control rod worth, sodium void worth, masses for major heavy nuclei, radial and axial power distribution at high burnup are studied. The results show that the reactor breeds fissile isotopes with a conversion ratio of 0.994 at fuel burnup 70 (GWd/T), and minor actinides are buildup inside the reactor core. The study aims to check the efficiency of the model on the calculation of the neutronic parameters of the core at high burnup.

Preliminary analysis of a zirconium hydride moderated megawatt heat pipe reactor

At present, megawatt-class Heat-Pipe Reactors (HPRs) are under-investigated widely as a candidate for robust, self-contained, and long-term power for special purpose applications. In this work, zirconium hydride is introduced to design a moderated megawatt-class HPR core. The influence of moderators on core parameters is analyzed. According to the calculation, although the use of a moderator will lead to higher burnup reactivity loss, the reactivity decreasing caused by the reduction of fuel can be offset by the proper moderator. The introduction of moderator has no significant influence on radial and axial power distribution and power peak factor. Meanwhile, the moderation of core will decrease the maximum neutron flux as well as the reactivity worth of control devices. The moderated core possesses a stronger doppler effect while the thermal expansion reactivity feedback of monolith and fuel decreases. Results also show that the introduction of moderators is beneficial to reducing the thickness and weight of the shield. Preliminary thermal analysis is conducted to verify the feasibility, the results show that zirconium hydride can stay stable in the moderated megawatt heat pipe reactor, and the yttrium hydride shows better thermal stability.

A 3-D Neutron Distribution Reconstruction Method Based on the Off-Situ Measurement for Reactor

Understanding the core neutron distribution is an indispensable condition to ensure the safe operation and effective burn-up of reactor fuel. It has become one of the new ways for reactor monitoring to use ex-core monitoring signals to calculate the in-core information recently. In this paper, an FCN-based architecture was implemented, where the 2D and 3D fully convolutional neural networks were used to reconstruct the space-and energy-dependent 3D neutron distribution of the reactor core through the off-situ measurement directly. The new method has been validated in the H. B. Robinson reactor. The differences between reconstruction results of neutron flux distribution and MC calculation were mostly less than 2%. The average relative deviation of reconstructing the neutron density kept in 5%. It was demonstrated that the provided method was applicable to monitor the reactor core theoretically and would provide strong support for the subsequent research on the real-time monitoring for an operating reactor, safety and operational testing for new reactor designs, and evaluation of source term information in a nuclear emergency.