Fissile Content Research Articles

Molten Uranium Breeder Reactor (MUBR) and Its Development Steps

Background The Molten Uranium Breeder Reactor (MUBR) is a radical new mixed-energy spectrum breed and burn reactor concept. The MUBR is fueled with molten uranium metal fuel in large fuel tubes instead of thin fuel rods, and is cooled by circulating the molten fuel through a heat exchanger. The purpose of this research is to find MUBR configurations with a SCALE (RSICC request/license 203869) burnup at least 10 times greater than the initial fissile content. Methods A proprietary computer program uses parameters to generate MCNP (RSICC request/license 176034) or SCALE input files and initiate MCNP or SCALE burn simulations and other analysis. This allows relatively fast comparison of different parameters to optimize the configuration. Results MUBR configurations have been found which have SCALE burn simulations through 120 years of fuel life with a burnup of 35% of the initial fuel mass when the initial fuel was Low Enriched Uranium (LEU) (3% U-235). Conclusions These results suggest that if it can be constructed and operated reliably for a long time, the MUBR would be a true breed and burn reactor. Compared to Light Water Reactors (LWR) this would provide many advantages for the nuclear industry and the world such as: reactor operation without shutdowns for refueling or fuel manipulation, no need for special fuel, ten times as much energy per ton of fuel, one tenth as much Used Nuclear Fuel (UNF) per megawatt hour of electricity, no UNF storage required every two years, increased energy security, and reduced nuclear proliferation risk. While development of the MUBR concept from a concept to a prototype reactor will be costly, the advantages suggest that the concept merits further study.

Demonstrating a Quantitative and Systematic Approach to Reducing Excess Conservativism in Nuclear Criticality Safety Analyses

Although reducing conservatism would alleviate unnecessary constraints in processing, storage, transportation, and disposal of nuclear materials, excessively conservative approaches are still utilized in many safety analyses. Criticality safety limits are put in place to reduce the likelihood of having a nuclear criticality accident to a value that is deemed incredible but often utilize parameters that are conservative to the point of becoming incredible themselves. The analyses that determine criticality limits are supposed to be based on credible instead of incredible events and circumstance, highlighting the need to be able to distinguish between what is in the realm of possibility and what is not. This paper provides a quantitative approach for reducing unrealistically conservative parameters by recalculating limiting factors in a state that deviates from the worst-case scenario and assigning probability distributions to these systematic deviations. This provides a technical basis for replacing excessively conservative values with something that is both objective and reasonably bounding, which may be systematically utilized in any criticality safety analyses. The assumption of “perfect sphericity” in the TRUPACT-II package’s fissile contents model was used as an example case to demonstrate the proposed approach for replacing qualitative reductions to conservatism with quantitative reductions. Through a series of Monte Carlo calculations and statistical analyses, it was shown that conservative deviations from sphericity will provide lower keff values, where the magnitude and impact of this deviation is system specific. The statistical significance from applying probabilistic conservatism will be dependent on the chosen κ value and integration limit for the exponential distribution, as it varies the degree of conservatism applied to any parameter of interest. This approach is not limited to geometric assumptions and may be applied to a variety of conservative parameters. In an effort to move toward a standard method for reducing conservatism, this objective approach may be used in lieu of or in conjunction with subjective methods for relaxing constraints.

Interrogating a Mixed Actinide Basket Using High-Resolution γ-Ray Spectrometry: A Nuclear Forensic Perspective on Possible Smuggling Scenarios.

The total fissile content in seized nuclear materials is of immense importance and needs to be estimated with reasonable accuracy as a part of nuclear forensics for early decision-making in legal proceedings. High-resolution γ-ray spectrometry (HRGRS), because of its nondestructive nature, is a powerful tool for the assay of such samples to reach a quick "on-site" decision on the severity, intended use, and associated radiological threat. If the seized package contains fissile isotopes of more than one actinide in a multicompartmental heterogeneous mixture, analogous to the most likely scenario of a "smuggled mixed actinide basket", its "on-site" quantification can be extremely challenging. This makes up an increasing share of the absolute HRGRS in nuclear forensics and demands for fundamentally new approaches. In the present work, the challenges associated with varying attenuation experienced by γ-rays of different actinides at different subcontainments of the heterogeneous sample matrix have been addressed by an iterative efficiency transfer approach from "point" to "extended" source for individual actinides and demonstrated for the assay of four mock-up samples and a legacy packet, mimicking seized packages containing nuclear materials. An absolute isotopic inventory of the fissile and other radioisotopes has been obtained within <10% along with the assay of total U and Pu within <3% of the expected values with measurement uncertainty <10% for the majority. The present approach has a good potential for "on-site" nuclear forensics in nuclear smuggling scenarios and also can be adapted easily for a wide variety of other applications.

In-air and in-water performance comparison of Passive Gamma Emission Tomography with activated Co-60 rods

A first-of-a-kind geological repository for spent nuclear fuel is being built in Finland and will soon start operations. To make sure all nuclear material stays in peaceful use, the fuel is measured with two complementary non-destructive methods to verify the integrity and the fissile content of the fuel prior to disposal. For pin-wise identification of active fuel material, a Passive Gamma Emission Tomography (PGET) device is used. Gamma radiation emitted by the fuel is assayed from 360 angles around the assembly with highly collimated CdZnTe detectors, and a 2D cross-sectional image is reconstructed from the data. At the encapsulation plant in Finland, there will be the possibility to measure in air. Since the performance of the method has only been studied in water, measurements with mock-up fuel were conducted at the Atominstitut in Vienna, Austria. Four different arrangements of activated Co-60 rods, steel rods and empty positions were investigated both in air and in water to confirm the functionality of the method. The measurement medium was not observed to affect the ability of the method to distinguish modified rod positions from filled rod positions. More extended conclusions about the method performance with real spent nuclear fuel cannot be drawn from the mock-up studies, since the gamma energies, activities, material attenuations and assembly dimensions are different, but full-scale measurements with spent nuclear fuel are planned for 2023.

Annular fuel for VVER-1000 reactor: Moderation impact on neutronic behavior

The impacts of moderation on the neutronic behavior of the VVER-1000 reactor assembly have been investigated by using annular fuel and compared with the conventional one. The study revealed that annular fuel shows a softer neutron spectrum, a better control rod worth and 16% less fissile content compared to the conventional fuel. After 60 MWd/kg burnup, annular fuel contains 6.45% and 255.41% more Pu-239 and Pu-241 respectively than the conventional fuel which can be a valuable source of fissile materials for the MOX fuel. Additionally, annular fuel shows a higher degree of symmetry in power peaking factor (PPF), higher scattering rate, 3.41% less tritium production at the beginning of cycle, and better shutdown margin without significant changes in axial & radial flux. From the neutronics point of view, annular fuel can be a propitious replacement of conventional fuel for the VVER-1000 reactor.

Implementation of in-rod axially heterogeneous thorium-uranium fuel in a typical PWR

Given the potential of the thorium fuel cycle and the predominance of PWRs in today’s commercial fleet of nuclear power plants, further investigations on the introduction of thorium-based fuel in open-cycle reactors such as PWRs are warranted. The benefits of microheterogeneous thoria-urania fuel (e.g. increased burnup / fuel cycle length, reduced radiotoxicity and increased stability of spent fuel, proliferation advantages due, feasibility of converting UO2-based fuel manufacturing plants to ThO2-UO2), are diminished by fuel manufacturing costs, the need to denature thorium sections of the fuel, thermo-hydraulic design requirements, higher fission gas release and cladding temperature gradients.This paper presents results of the neutronic performance of homogeneous and multiple heterogeneous fuel pin cell designs and explores the fundamental reasons and reaction balances between the different fuel models. The work utilises fuel models with variations in the length and composition of thorium and uranium sections. We compare the various urania-thoria fuel models with a reference PWR (North-Anna 1) with similar fissile content. We postulate a number of fuel design criteria, to appraise fuel models in trade-off studies, including improved burnup and cycle length, limitations on fuel enrichment grade and excess reactivity during burnup (typically at beginning of cycle), safety requirements such as limitations on total burnup, power peaking and feedback mechanisms. In addition, criteria are set for proliferation resistance and radiotoxicity.Verification studies were performed on fuel with homogeneous and axially heterogeneous fuel cell pins, followed by initial assembly and full-core analyses, to affirm confidence in the models employed. Our results confirm significant conversion of thorium to 233U, resulting in an improved burnup and fuel cycle length. Shut down margins and neutron economy are improved by significantly reducing beginning of cycle excess reactivity and hence reducing the need for burnable and soluble neutron absorbers. The conversion of thorium to 233U then leads to an increase in reactivity later in the cycle. To address proliferation concerns, fuel models with denatured thorium sections are introduced.We conclude that among the well-performing fuel models, fuel comprising heterogeneous pins repeating 3 cm thorium and 1 cm uranium sections, with the uranium region enriched to 20 at% 235U, and the thorium region denatured to a specific degree, performed best in terms of the fuel criteria considered. The promising results for the fuel pin cell models were confirmed with initial assembly and full core models: Burnup improvement rendered by the 3Th1U fuel models is 25.6% / 22.3% / 17.9% for the pin cell / assembly / full core models respectively, in spite of employing denaturing of thorium sections, while also fulfilling the other stated fuel criteria.

Analysis of the Effect of Central Gap Size on the Temperature Profile and Fissile Content in the Annular Nuclear Fuel Rod

The perturbation in temperature profile and fissile content due to variation in the central gap of an isolated annular cylindrical fuel rod of a pressurized water reactor was studied by analytical calculation. Four different models namely UO2+Zircaloy-4+He, UO2+ Zr-1%Nb +He, MOX+Zircaloy-4+He, and MOX+ Zr-1%Nb +He were considered. The radial temperature profile was generated for different ratios (α) of outer to inner fuel radius. Lower fuel temperature was observed for small values of α and vice versa. The peak fuel temperature and temperature drop across the fuel pellets were calculated. Models of MOX fuel showed higher peak fuel temperature and large temperature drop than the models of UO2 for the same fuel cladding. Zr-1%Nb cladding results in a slightly higher fuel temperature than Zircaloy-4 for the same fuel composition. The faster changes of these parameters with α were found for UO2 than MOX fuel. The change in fissile loading with α was also studied and a sharp increase is observed if exceeds α=2.50. DUJASE Vol. 6 (2) 72-76, 2021 (July)

Nuclear proliferation resistance assessment of fuel cycles closed with complete co-processing of spent fuel

Purex (Plutonium Uranium Recovery by Extraction) is the standard reprocessing method used for the recovery of Uranium (U) and Plutonium (Pu) in spent fuel. Products of the standard reprocessing of spent fuel discharged from a typical LWR are separated and highly pure streams of U and Pu with fissile contents of roughly 0.85 wt.% (U-235 in U) and 70 wt.% (Pu-239+Pu-241 in Pu) respectively. Because of the economic and technological conditions and safeguard concerns, alternative methods or modifications for standard reprocessing including complete co-processing have been proposed. Complete co-processing is based on the Purex process and is an easier way of recovering U and Pu from spent fuel compared to standard reprocessing. In addition, since the U and Pu recovered together from spent fuel, it does not yield pure Pu and it is advantageous with regard to safeguards. The product obtained from complete co-processing of spent fuel is a U + Pu mixture with a total fissile content of around 1.5 wt% (U-235+Pu-239+Pu-241 in U + Pu), which is not suitable for direct use in LWRs and special approaches are required for recycling it. This study focuses on the assessment of nuclear proliferation resistance of fuel cycles closed with complete co-processing and different recycling scenarios for U + Pu product. To assess the proliferation resistance, intrinsic proliferation barriers including the amount of material to be diverted for significant quantity (SQ) of Pu, isotopic content of Pu, and radiation barrier of diverted material are compared for spent fuels generated in the cycles for one GWe-yr electricity production. To assess the impact of burnup on the proliferation resistance of cycles, all calculations are performed for the burnup values 33 GWd/tHM, 40 GWd/tHM, and 50 GWd/tHM. Results showed that the closed fuel cycle in which the U + Pu product from the complete co-processing is blended with 10 wt.% enriched uranium and recycled as mixed oxide fuel possesses the highest proliferation resistance. Besides, intrinsic nuclear proliferation barriers of all the fuel cycles investigated are positively affected by an increase in burnup.

Neutronic analysis of CANDU fuel bundle with reprocessed fuel

Fuel economy and novel fuel cycles for nuclear industry is a wide subject that required a massive effort along the decades. The reprocessing methods has shown a great potential to reach this goal to reduce the environmental issues related to burnt fuel disposal in nuclear resources utilization frame which is the primary subject. In second to reduce the mining and milling disadvantages which has minor effect if compared with the first. Besides the advantages reprocessing wastes should be considered. Thus, advanced nuclear fuel cycles has been proposed and one of the interest for CANDU technology is the communication between CANDU and LWR fuel cycles. In this concept, for short DUPIC/OREOX method is one of the most promising due cost and lower residual products and retain non-proliferation features. Since the spent PWR fuel has a higher fissile content than natural uranium, the DUPIC fuel results in a greater amount of fissile nuclides in fuel matrix. The objective of this paper is to state a roadmap to keep neutronics of the reprocessed fuel bundle as closer as possible to natural uranium fuel bundle to achieve a reliable scenario for further CANDU 6 reactor core fueling scheme. An evaluation of neutron absorber materials (depleted uranium, thorium, dysprosium and hafnium) was presented in sense of reactivity reduction due reprocessed fuel attributes. In the systems assessment the use of hafnium had shown to be suitable relating to system criticality. The SCALE 6.0 code was used for modeling and analysis.

Sensitivity of the active neutron coincidence collar response during simulated and experimental fresh fuel assay

Verification of the fissile (and fertile) content in fresh nuclear fuel assemblies is conducted by the IAEA to enforce the Nuclear Non-Proliferation Treaty using the UNCL (Uranium Neutron Collar — Light Water Reactor Fuel). The UNCL uses an uncorrelated AmLi neutron source to interrogate the fuel, producing a signal of coincident fission neutrons (Doubles rate) used to assay 235U content of the fuel. The cost of producing calibration assemblies and limited availability of diverse commercial assemblies at any one time historically restricted the ability to explore the full parameter space experimentally. Monte Carlo simulations can overcome this, but introduce additional sources of uncertainty. Here the sensitivity of simulations and measurements to various parameters is assessed for a reference 16×16 PWR assembly of uniform 3.2% enrichment. Uncertainty contributions in this evaluation include: simulated AmLi neutron source spectrum, AmLi neutron emission rate, AmLi anisotropicity, high density polyethylene (HDPE) density, the precise position of the fuel assembly within the detector, and experimentally the statistical uncertainty. The overall total systematic uncertainty estimate for the simulation of the absolute/relative Doubles rates responses are estimated to be approximately 2.0%/1.5%, and for experimental measurements systematic uncertainty reduces to 1.1%. This analysis supports further work using the relative Doubles rates in place of measurements for updating and extending the UNCL analysis methodology as systematic uncertainty is reasonably small.

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions: Results of BZ-3 and BZ-4 tests

Fission gas release from high-burnup mixed-oxide (MOX) fuel pellet with Pu fissile contents of 4.1% during reactivity-initiated accident (RIA) was estimated based on radial Xe concentration profiles in fuel pellet using scanning electron microscopy/electron probe microanalysis (SEM/EPMA) before and after pulse irradiation tests called BZ-3 and BZ-4. The tests were conducted at the Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Agency (JAEA). The depression of Xe concentration in the Pu-spot and the crack formation in the peripheral region of fuel pellet were observed after BZ-3 and BZ-4. This kind of cracks was considered to promote the release of the fission gas retained in the Pu-spot. Based on the EPMA results and the effect of the crack formation, the amounts of the release of the gas retained in Pu-spot was estimated to be 0.13 and 0.04 in BZ-3 and BZ-4 tests, respectively. Moreover, taking account of the release of the gas accumulated in the grain boundary in Pu-spot-excluded region and the release of the gas dissolved in the matrix, the total amount of fission gas release was evaluated as 0.33–0.43 and 0.15–0.27 in BZ-3 and BZ-4 tests, respectively. Comparing the results of the fission gas release with the results of rod-puncture test, it was suggested that fission gas release from not only the Pu-spot but also the Pu-spot-excluded region.

Passive neutron albedo reactivity measurements of spent nuclear fuel

The upcoming disposal of spent nuclear fuel in Finland creates new challenges for nuclear safeguards. Part of the national safeguards concept for geological repositories, developed by STUK — Radiation and Nuclear Safety Authority, is non-destructive assay (NDA) verification of all fuel items before disposal. The proposed verification system is a combination of PGET (Passive Gamma Emission Tomography), PNAR (Passive Neutron Albedo Reactivity) and weight measuring NDA-instruments. PGET takes a pin-level image of the fission products inside of a fuel assembly and PNAR verifies the multiplication of the assembly, a quantity that correlates with the fissile content. PGET is approved by IAEA (International Atomic Energy Agency) for safeguards measurements, but the feasibility of PNAR has not yet been established. A first of its kind PNAR prototype instrument was built in a collaboration coordinated by STUK. This paper concludes the results of the first measurements of spent BWR (Boiling Water Reactor) nuclear fuel with the prototype in July 2019. Based on the measurements, the ability of the PNAR instrument to detect the presence of fissile material in a repeatable manner in a reasonable amount of time was demonstrated. Furthermore, the instrument was able to detect differences in multiplication between partially and fully spent fuel assemblies, and axial differences in multiplication within a single assembly.

Optimization of AHWR core design: neutronics and thermal-hydraulics analysis and calculations

Advanced Heavy Water Reactor(AHWR) is being designed by BARC for thorium utilisation and demonstrate all aspects of the thorium fuel cycle. The AHWR is a vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The equilibrium fuel cycle of AHWR is being optimised for a discharge burn-up of 40 GWd/Te with Plutonium as makeup fuel in a closed fuel cycle. The 233U content in the Uranium and fissile Plutonium content in the Plutonium have been assumed to be about 78% and 75% respectively. One of the important passive safety features of AHWR is heat removal through natural circulation which is governed by a strong neutronic and thermal hydraulic coupling. The neutronics design simulations were done for average coolant condition. In order to study the effect of natural circulation and its consequences on the coolant behaviour and the resulting power distribution, the neutronics and thermal–hydraulic coupled calculations were carried out. The neutronics calculations code (FEMFOL) and thermal–hydraulic calculations code (ARTHA) were externally coupled to estimate the coolant density distribution and the critical heat flux ratio (CHFR) for the fuel. This paper gives the details of the critical heat flux ratio estimation and the effect of density distribution on the core power distribution at some selected operating phases during core follow-up of AHWR.

A Novel Approach to Realistic Conservatism in Nuclear Criticality Safety Analysis

The standard approach in nuclear criticality safety analysis is to rely quite heavily—and in some cases exclusively—on passive controls, such as assuming all worst-case conditions are by default attained. This means assumptions are made such as no poison, optimum moderation, and pure fissile actinide content at the maximum mass with optimum full reflection. What is clearly attainable is something less than any of these extremal conditions, but how can one rely on a limit based on assuming less than the worst case without some controls ensuring those assumptions are not challenged? This technical note discusses various options for approaching a defendable realistic technical basis for safety analysis by associating probabilities with conservative assumptions.

Evaluation of hot channel factor for sodium-cooled fast reactors with multi-physics toolkit

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Comparison of the single and double count using MCNP6 and ONMS Geant4 software

The quantification and detection of fissile material are valuable for safeguards, nuclear security, and fuel management. Non-Destructive Assay (NDA) (i.e. using neutron and gamma measurements) provide reliable means to detect and estimate the fissile content (i.e. mass) in a vast range of material in different forms found across various facilities in the nuclear fuel cycle. Neutron measurements are used either for total counting (i.e. pure metal fissile samples) or coincidences/multiplicity counting (i.e. impure fissile samples). These various flavors of neutron measurement are routinely used to quantify fissile material mass and inventory lists. Monte Carlo codes have evolved considerably and are used routinely for modeling and simulation of diverse physics. MCNP6 (Monte Carlo N Particle) and ONMS-Geant4 (Open Neutron Multiplicity Software) are used in this work to generate single and double count (i.e. based on the accidental and real plus accidental histograms) for two fresh fuel sources with the assumption that the neutron emission process is solely due to spontaneous fission. Previous work has shown that for these fresh fuel sources, about 93% of the neutrons emitted is due to spontaneous fission. ONMS was designed to provide an open-source platform that will facilitate researchers in the nuclear community to perform neutron counting simulations. ONMS has been shown to give good results for detectors with high efficiencies, strong (α, n) sources, strong spontaneous fission sources and sources baring plutonium. Detailed model simulations result from ONMS and MCNP6 are compared for a weak californium source and uranium-bearing fresh fuel assembly using a detector set up with relatively low efficiency. The detector response is characterized to compare the single and double counts obtained from both code packages. For the weak californium source, the neutron triplets are also evaluated from both code packages.

Estimation of fissile material content in irradiated In-Pile Sections using neutron coincidence counters

A set of In-Pile Sections (IPS) has been irradiated in the BR2 reactor at SCK•CEN in Belgium during the 1970’s and 1980’s. The primary goal of the IPS was to replicate the thermo-hydraulic loop of a sodium-cooled fast reactor in order to study severe accident scenarios. The top part of the IPS contained the sodium-cooled loop whereas the lower part contained the fuel element. Due to the experimental conditions, the rupture of the fuel pins contained in the IPS occurred and fuel fragments may have been deposited in the rest of the IPS loop. The part of the IPS containing the fuel pins has been cut from the rest of the IPS and underwent post-irradiation examinations at specialized EU laboratories, while the top parts remained stored at SCK•CEN. To prepare for future transport, dismantling and conditioning, a reliable estimation of the total fissile content in the stored parts of the IPS is indispensable. In this framework, two IPS were measured with a Canberra WM3400 neutron coincidence counter with customized electronics. The measurements of the IPS were challenging due their length (roughly 6 m) and intense gamma-ray radiation background. For each IPS an axial scan was carried out with a series of short measurements (600-700 s each) recording the Totals rate and Reals rate. Based on the results of the axial scans, measurements with longer measurement time were conducted for the axial positions with the larger values of Reals rates. A system of equations was then established to quantify the 240Pu content in the different sections of the IPS from the Reals rates in each measurement position and account for cross-talk between the neutron emission associated to the different sections. A set of Monte Carlo simulations was carried out to estimate the probability to record a Real count in the detector due to spontaneous fission events occurring in a given section of the IPS. The 240Pu content in each section of the IPS was calculated by combining the measured Reals rates and the detection probabilities calculated with the simulations. The total fissile content in the IPS was then determined with scaling factors based on burnup calculations for the irradiated fuel assemblies in the IPS. The results indicate that both IPS measured with the neutron coincidence counters have a fissile content lower than the limit for transport. It is expected that the envisaged segmentation of the IPS in shorter sections required to fit into 200L drums will provide an additional safety margin on this limit.

Basic design parameter optimization on water cooled thorium breeder reactor

Basic study on optimization of heavy water cooled thorium breeding reactor has been investigated to estimate the feasible design region of breeding and negative void reactivity coefficient. Those parametric surveys are moderator to fuel ratio, fuel pin diameter, fuel pellet power density, and void faction. A fuel pin diameter of 14.5 mm shows the optimum feasible fuel breeding and less required U-233 fissile content. The systems require the enrichment less than 8% and it gives some possible area of breeding. The effect of moderator is essential to make the negative void reactivity. Void fraction effect is still good enough to make more negative void reactivity, however, for no more heavy water as moderator, it obtains less negative void reactivity or becomes slightly positive. It has been shown that several significant parameters affect to the feasibility breeding windows and potential negative void reactivity based on heavy water-cooled thorium reactor.

Stand-off nuclear reactor monitoring with neutron detectors for safeguards and non-proliferation applications

Safeguards measures are employed at nuclear reactor facilities worldwide, to ensure that nuclear material is not diverted from peaceful uses. Typical safeguards measures involve periodic inspections, off-line verification and video surveillance of fuel cycle activities. Real-time verification of the fissile contents via stand-off monitoring can enhance continuity of knowledge for non-traditional reactor types, including research reactors and small modular reactors. Here we demonstrate the feasibility of using large-area neutron detectors for monitoring nuclear reactors at stand-off distances up to 100 m outside reactor shielding, as a potential reactor safeguards tool. Since the neutron yield per unit reactor power depends upon the isotopic composition of the reactor core, declared changes in fissile composition can be verified without accessing the core. The supporting results of experiments conducted at the National Research Universal reactor in Canada, are presented.

Nuclear isotopic fissile content assay

Isotopic fissile content assay was performed in a designed lead slowing down spectrometer (LSDS) system. In the system, the fission measurement was performed for 11 – 20 wt% of Pu239 with 1 wt% of Pu241 and U235. The self-shielding factor was applied to the measured fission signal. The fissile assay was performed on U235, Pu239, and Pu241. The content assay of an unknown fissile sample was also performed on U235, Pu239, and Pu241. The procedure of an unknown sample assay was setup. In several simulations, the content of Pu239 was accurately obtained with ∼1% error in the rod base. For the unknown fissile assay, the result was very close to the actual content of Pu239, ∼1% error. The LSDS provided very powerful result for the fissile assay in the mixture. Therefore, the LSDS is a very promising technique in isotopic fissile content assays, especially for Pu239 in uranium and plutonium. Precise fissile material content should be known for reactor safety in fuel cycles and international safeguards on nuclear fuel cycle development.