Fast Neutron Spectrum Research Articles

Calculation verification of the utilization of LR-0 for reference neutron spectra

Abstract Well-defined neutron spectrum is crucial for calibration and testing of detectors for spectrometry and dosimetry purposes. As a possible source of neutrons nuclear reactors can be utilized. In reactor core most of the neutrons are originated from fission and neutron spectra is usually some form of moderated spectra of fast neutrons. The reactor LR-0 is an experimental light-water zero-power pool-type reactor originally designed for research of the VVER type reactor cores, spent-fuel storage lattices and benchmark experiments. The main reactor feature that influences the performance of experiments is the flexible arrangement of the core. Special types of the possible core arrangements on the reactor LR-0 can provide different neutron spectra in special experimental channels. These neutron spectra are modified by inserting different materials around the channel and whole core is driven by standard fuel assemblies. Fast, epithermal or thermal spectra can be simulated using graphite, H 2 O, D 2 O insertions, air, Cd foils or fuel with different enrichment.

MYRRHA: A multipurpose nuclear research facility

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback.As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

Impact of Thorium Capture Cross Section Uncertainty on the Thorium Utilized ADS Reactivity Calculation

Recently, the researches on fast neutron spectrum system utilized thorium fuel are widely conducted. However, the recent thorium cross section libraries are limited compared to uranium cross section libraries. The impact of thorium cross section uncertainty on thorium fuel utilized accelerator driven system (ADS) reactivity calculation is estimated in this study. The uncertainty of thekeffcaused by232Th capture cross section of JENDL-4.0 is about 1.3%. The uncertainty of JENDL-4.0 is needed to be enhanced to provide more reliable results on reactivity calculation for the fast system. The impact of uncertainty of 232Th capture cross section of ENDF/B-VII is small (0.1%). Therefore, it will cause no significant impact of the thorium cross section library on the thorium utilized ADS design calculation.

Compilation of Existing Neutron Screen Technology

The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs) by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.

Towards the thorium fuel cycle with molten salt fast reactors

There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. It has been recognized as a long term alternative to solid-fueled fast neutron systems with a unique potential (large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction etc.) and is thus one of the reference reactors of the Generation IV International Forum. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions, so that the MSFR concept may use as initial fissile load, 233U or enriched uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management via stockpile incineration in MSRs.

Reconstruction of the fast neutron spectrum using deformation of a reference spectrum represented as a Legendre polynomial expansion

This paper proposes a method for reconstructing the fast neutron spectra in the range 0.2 MeV to 100 MeV based on modified methods of the effective threshold cross sections and the deformation of a reference spectrum using the Legendre polynomial expansion of a deforming function. In contrast to the deformation method traditionally used to determine the desired differential spectrum φ(E), we propose the use of a deforming function as a Legendre polynomial expansion in order to determine the integrated spectrum F(E). Since F(E) is a monotonic decreasing function that tends to zero when approaching the maximum spectral energy, it can be represented by a shorter series; i.e., the number M of the sought series coefficients am will not be large. The developed program was used to calculate fast neutron spectra in a QUINTA nuclear setup that was irradiated with 1–8 GeV deuterons of Nuclotron-M at the Laboratory of High Energies of the Joint Institute for Nuclear Research.

TERRAPOWER, LLC TRAVELING WAVE REACTOR DEVELOPMENT PROGRAM OVERVIEW

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ∼30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

Measurement of fast neutron spectra from the interaction of 20 MeV protons with thick Be and C targets using CR-39 detector

Fast neutron spectral yield from the interaction of 20MeV protons with thick Be and C targets are measured using CR-39 detectors. An image analysing program (autoTRAK_n) dedicated for neutron spectrum generation and dose estimation is developed based on the analysis of greylevel variations within each track and is used in the present study. The most important parameters for neutron spectrometry, i.e., track depth (length) and angle of all neutron-induced recoil tracks are determined using autoTRAK_n. The methodology is tested successfully to reproduce the spectra of some standard neutron sources which was reported earlier. In this study, autoTRAK_n is applied to generate neutron spectra from the above mentioned reactions, i.e., 9Be(p, n), 12C(p, n) and the neutron yields (total number of neutrons per projectile) for both the reactions are determined by folding the track density with the detector response. The dose equivalents and H∗(10)-to-fluence ratios are also estimated. All these quantities are automatically obtained from the program. The neutron yield and the dose equivalent for 9Be(p, n) reaction are found to be about 4 times higher than that for 12C(p, n) reaction. The present methodology for neutron spectrum generation is found to be simple and effective and does not involve the complex spectrum unfolding procedures.

Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit

Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing.To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor+reprocessing unit) by a system of equations that can be numerically solved.Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit rate for the reprocessing is 2.5l of fuel salt a day, which means that the fuel should be reprocessed within 7000days approximately if there is a specific way to control the redox potential of the salt. Finally, a last part of this paper analyzes the impact of chemical parameter uncertainties on the reprocessing performance.

Fuel Salt for the Molten-Salt Reactor

The solubility of Th, U, Ce, La, and Pr in melted salt 45LiF–12NaF–43KF, which supports a fast neutron spectrum for the molten-salt reactor, is found experimentally to be high in the interval 550–700°C. On the basis of an analysis of the physicochemical and nuclear-physical properties of different fluoride salts, it is proposed that the eutectic 46.5LiF–11.5NaF–42KF be used as the fuel salt for a molten-salt fast reactor with a uranium-plutonium fuel cycle.

Fast neutron spectrum unfolding of a TRIGA Mark II reactor and measurement of spectrum-averaged cross sections: integral tests of differential cross sections of neutron threshold reactions

Summary The spectrum of fast neutrons having energies from 0.5 to 20 MeV in the core of the 3MW TRIGA Mark II reactor at Savar, Dhaka, Bangladesh, was unfolded by activating several metal foils to induce threshold nuclear reactions covering the whole spectrum, and then doing necessary iterative calculations utilizing the activation results and the code SULSA. The analysed shape of the spectrum in the TRIGA core was found to be similar to that of the pure 235U-fission spectrum, except for the energies between 0.5 and 1.5 MeV, where it was slightly higher than the fission spectrum. Spectrum-averaged cross sections were determined by integral measurements. The integral values measured in this work were compared with the recommended values for a pure fission spectrum as well as with the integrated data deduced from measured and evaluated excitation functions of a few reactions given in some data files. The good agreement between integral measurements and integrated data in case of well-investigated reactions shows that the fast neutron field at the TRIGA Mark II reactor can be used for validation of evaluated data of neutron threshold reactions.

High enriched to low enriched fuel conversion in YALINA Booster facility

The YALINA Booster zero power facility is a subcritical assembly located in Minsk (Belarus). This assembly has the special feature of operating on fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in polyethylene matrix. This study discusses the high enriched to low enriched fuel conversion. In order to increase the neutron multiplication of the assembly loaded with low (21%) enriched fuel in the fast zone, the number of fuel rods in the thermal zone cannot be augmented. Consequently, the effective multiplication factor of the configuration with 21% enriched uranium fuel in the fast zone has been enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located in-between the fast and the thermal zones, to form a circular rather than a square arrangement. The MCNP computer simulation results obtained with the circular arrangement of the absorber rods are in good agreement with the experimental data.

Coupled 3-D Neutronics/Thermal-Hydraulics Optimization Study for Improving the Response of a 3600 MW(Thermal) SFR Core to an Unprotected Loss-of-Flow Accident

The sodium-cooled fast reactor (SFR), as a fast-neutron spectrum system, is characterized by several performance advantages. In particular, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. However, the SFR has one dominating neutronics drawback, namely, there is generally a positive reactivity effect when there is voiding of the sodium coolant in the core. Furthermore, this effect becomes even stronger in the equilibrium closed fuel cycle. Considering that in a hypothetical SFR unprotected loss-of-flow (ULOF) accident scenario, i.e., flow rundown without SCRAM, sodium boiling can be anticipated to occur, it is crucial to assess the corresponding impact of the positive sodium void effect.An optimization study for improving the safety characteristics of a large [3600-MW(thermal)] SFR has currently been conducted in the above context. The dynamic core response to a reference ULOF scenario is investigated with the use of a coupled three-dimensional neutronics/thermal-hydraulics PARCS/TRACE model. The starting point of the study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the sodium void effect, the core has been modified by introducing an upper sodium plenum, along with a boron layer above it. Furthermore, the original core height-to-diameter ratio is reduced. In comparison to the reference ESFR core behavior, certain improvements are achieved, thanks to the static neutronics optimization carried out. However, these changes are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting during the considered ULOF event.Thermal-hydraulics optimization has thus been considered necessary, in order to (a) prevent sodium flow blockage in the fuel channel and (b) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. The corresponding measures taken are (a) the introduction of an innovative wrapper design, which features small openings in each side surface of the fuel assembly, and (b) replacement of the original upper sodium plenum by an extended fission gas plenum. Following implementation of these thermal-hydraulics-related design changes, one arrives at a final configuration of the SFR core, in which, for the selected accident scenario, a new “steady state” involving stable sodium boiling is found to be achievable, with melting of neither cladding nor fuel. Such a satisfactory behavior has been confirmed not only for the beginning-of-life state of the core but also for the equilibrium closed fuel cycle.

Investigation of the MSFR core physics and fuel cycle characteristics

The adoption of Th fuel in fast reactors is being reconsidered due to the potential favorable impact on actinide waste management and resource availability. A closed Th cycle leads to an actinide inventory with lower radiotoxicity and heat load for the first several thousands of years. Due to the typically low TRansUranic (TRU) Conversion Ratio (CR), Th can also be advantageous to expedite the consumption of legacy TRU. One of the main obstacles to the implementation of Th is the highly radioactive recycled fuel which requires remote handling under heavy shielding, inevitably penalizing economics and challenging conventional pin-based fuel manufacturing. From this perspective, the development of liquid-fuelled reactors, with Molten Salt Reactors regarded as the most promising, appears particularly attractive as fuel handling would be greatly simplified. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radiotoxicity generation and safety-related parameters. Similarly to most MSR concepts proposed in the past, the MSFR is based on the fluoride molten salt technology, but it features the novelty of a fast neutron spectrum. Calculations are performed using state-of-the-art equilibrium-cycle methodologies, i.e., the ERANOS-based EQL3D procedure developed at the Paul Scherrer Institut and extended to the simulation of the MSFR. Selected results have been benchmarked with the Monte Carlo code PSG2/SERPENT. These results have also been used for the assessment of a diffusion module based on the COMSOL multi-physics toolkit, which is the subject of current studies aimed at efficiently simulating the peculiar MSFR transient behavior.

Transmutation of high‐level nuclear waste by means of accelerator driven system

To be able to answer the worlds’ increasing demand for energy, nuclear energy must be part of the energy basket. The generation of nuclear energy produces, besides energy, also high‐level nuclear waste, which is nowadays for geological storage. Transmutation of the minor actinides and long‐lived fission products that arise from the reprocessing of the nuclear waste can reduce the radiological impact of these radioactive elements. Transmutation can be completed in an efficient way in fast neutron spectrum facilities. Both critical fast reactors and subcritical accelerator driven systems are potential candidates as dedicated transmutation systems. Nevertheless, an accelerator driven system operates in a flexible and safer manner even with a core loading containing a high amount of minor actinides leading to a high more‐efficient transmutation approach. WIREs Energy Environ 2014, 3:60–69. doi: 10.1002/wene.82This article is categorized under: Fossil Fuels > Science and Materials Fossil Fuels > Systems and Infrastructure Fossil Fuels > Climate and Environment

Transmutation of Nuclear Waste and the future MYRRHA Demonstrator

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at a short introduction into the field that has been characterized by a high collaborative activity during the last decade in Europe, in order to focus, in its later part, on the MYRRHA project as the European ADS technology demonstrator.

Measurement of neutron yield by 62 MeV proton beam on a thick Beryllium target

In the framework of research on IVth generation reactors and high intensity neutron sources a low-power prototype neutron amplifier was recently proposed by INFN. It is based on a low-energy, high current proton cyclotron, whose beam, impinging on a thick Beryllium converter, produces a fast neutron spectrum. The world database on the neutron yield from thick Beryllium target in the 70 MeV proton energy domain is rather scarce. The new measurement was performed at LNS, covering a wide angular range from 0 to 150 degrees and an almost complete neutron energy interval. In this contribution the preliminary data are discussed together with the proposed ADS facility.

Reactivity effect breakdown calculations with deterministic and stochastic perturbations analysis – JEFF-3.1.1 to JEFF3.2T1 (BRC-2009) actinides application

JEFF-3.1.1 is the reference nuclear data library in CEA for the design calculations of the next nuclear power plants. The validation of the new neutronics code systems is based on this library and changes in nuclear data should be looked at closely. Some new actinides evaluation files at high energies have been proposed by CEA/Bruyeres-le-Chatel in 2009 and have been integrated in JEFF3.2T1 test release. For the new release JEFF-3.2, CEA will build new evaluation files for the actinides, which should be a combination of the new evaluated data coming from BRC-2009 in the high energy range and improvements or new evaluations in the resolved and unresolved resonance range from CEA-Cadarache. To prepare the building of these new files, benchmarking the BRC-2009 library in comparison with the JEFF-3.1.1 library was very important. The crucial points to evaluate were the improvements in the continuum range and the discrepancies in the resonance range. The present work presents for a selected set of benchmarks the discrepancies in the effective multiplication factor obtained while using the JEFF-3.1.1 or JEFF-3.2T1 library with the deterministic code package ERANOS/PARIS and the stochastic code TRIPOLI-4. They have both been used to calculate cross section perturbations or other nuclear data perturbations when possible. This has permittted to identify the origin of the discrepancies in reactivity calculations. In addition, this work also shows the importance of cross section processing validation. Actually, some fast neutron spectrum calculations have led to opposite tendancies between the deterministic code package and the stochastic code. Some particular nuclear data (MT=5 in ENDF terminology) seem to be incompatible with the current MERGE or GECCO processing codes.

Multi-purpose fast neutron spectrum analyzer with real-time signal processing

Diagnostics of hot ion component of plasma on the products of fusion reactions is widely used on thermonuclear facilities. In case of employment of neutron spectrometers, based on organics scintillators, there is advanced technique developed to eliminate neutron pulses from gamma background—digital pulse shape discrimination. For every DPSD application it is necessary to use the fast (2–5ns) and precise (12bit) transient ADC unit with large amount of onboard memory for storing every digitized scintillation pulses during shot time. At present time the duration of hot thermonuclear plasma burning in large tokamaks approximate to 1min, and this requires very high onboard memory capacity (∼100GB). This paper describes a neutron spectrum analyzer with real-time DPSD algorithm, implemented to ADC unit. This approach saves about two orders of onboard memory capacity, gives the possibility of instant use of outcome to feedback systems. This analyzer was tested and calibrated with help of 60Co and 252Cf radiation sources, and deuterium neutron generator.

Design Study of Small Pb-Bi Cooled Non-Refueling Nuclear Power Reactors (SPINNORs)

Small Pb-Bi Cooled Non-refueling Nuclear Power Reactors(SPINNORs) are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for 20 years without on-site refuelling. In this study the results of neutronic design for 62.5-250MWt NPPs are discussed. The reactor cores are designed to have near zero (less then one effective delayed neutron fraction) burn-up reactivity swing during the burn-up period to avoid a possibility of super-prompt critical accident. The basic concept is that the reactor core is adjusted so that its internal conversion ratio is nearly one. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. In general higher fuel volume fraction is needed for smaller core.