Excess Reactivity Research Articles

The Method to Predict First Critical Core Loading for Nuclear Reactors

Nuclear power plants have an important role in carbon free electricity production in the world. One of the important steps of commissioning a nuclear power plant is the first core loading. This is also called approaching the criticality. Since the number of fuel elements for the criticality is not known, precautions must be taken to prevent safety incidents. Although the procedure is performed on-line such that the neutron counts are measured at each loading of fuel elements to calculate sub-critical multiplication and the number of fuel element to reach criticality were predicted, computer simulations can also be used. In this study, inverse sub-critical multiplication method was applied to Istanbul Technical University TRIGA Mark II research reactor first criticality in 1979 by using Monte Carlo simulation code MCNP6.2. Full 3-D model of the reactor was generated for calculations. Both results, experimental and simulation, showed that reactor became critical with 62 fuel elements. The core excess reactivity of 23.1 cents was predicted as 21.7 with the code. The simulation results are in good agreement with experimental results. The methodology and simulations can be used for power reactor analysis as well.

Use of erbium as a burnable absorber for the VVER reactor core life extension

The paper presents the results of a computational and theoretical analysis concerned with the use of erbium as a burnable absorber in VVER-type reactors. Partial refueling options for the reactor life extension to 18 and 24 months is considered, the refueling ratio being equal to three for the 18-month life and to two for the 24-month life. Erbium is expected to be present in all fuel elements in the FA with the same weight content. The influence of the erbium weight content on such neutronic characteristics of the reactor and fuel as burn-up, reactivity coefficients, residual volume of “liquid" control, and amounts of the liquid radioactive waste (LRW) formed was assessed. The calculations were performed using a simplified model of refueling without FA reshuffling. An infinite array of polycells consisting of FAs with different in-core times was considered. The escape of neutrons from the core was taken into account by selecting the critical value K∞ at the end of life. Erbium does not burn up in full for the lifetime which affects the fuel burn-up as compared with the liquid excessive reactivity compensation system. The reduction is 0.7% per 0.1% of the erbium weight load in the fuel elements. This, however, also reduces the maximum content of the boron absorber in the coolant and the LRW accumulation in the ratio of 5% per 0.1% of the erbium weight load. Erbium influences the spectral component of the coolant temperature reactivity coefficient which turns out to be negative even with its minor weight fraction in fuel elements, and a reduction in the boron absorber fraction leads to a positive value of the density reactivity coefficient. As a result, the overall coolant temperature reactivity coefficient has a negative value throughout the lifetime.

Neutronic analyses of americium burning U-free inert matrix fuels

Burnup characteristics of americium burning U-free inert matrix PuO2+ZrO2 (ROX), and PuO2+ThO2 (TOX) with reactor- and weapons-grade plutonium fueled LWR cell are analyzed in comparison with a conventional UO2 and MOX fueled LWR cells. In these fuels, 212–1062 kg/GWe of AmO2 are added to optimize an adequate amount for an effective burning of Am and Pu as well. It is found that a small amount of AmO2 act as a burnable poison in UO2 fuel to decrease the excess reactivity at BOL. Thermal neutron spectrum of ROX + AmO2 fuel is softer than the other fuels due to the lack of secondary source of fissile material, which accelerate the fission reaction rates of 242MAm, 242GAm, 243Am and capture reaction rates of 241Am, 243Am. As a result, the effective burning of Am and Pu isotopes especially 241Am and 239Pu are great in amount in ROX + AmO2 fuel even though the production of 242Cm, 244Cm, 238Pu from Am are nearly same in these fuels. In ROX + AmO2 fuel, the production of long-lived fission products (LLFPs) and 237Np are little. By analyzing the quality of spent fuels and safety characteristics, 650–750 kg/GWe of Am are proposed for ROX fuel for once through burning.

Preferred core conceptual design of pebble bed advanced high temperature reactor

Key parameters of Pebble Bed Advanced High Temperature Reactor (PB-AHTR) core are studied, and then the preferred core conceptual design is proposed. Both evacuating molten salt and injecting poison into coolant can be used as the auxiliary system of the second shutdown system in PB-AHTR. Compared with injecting poison into coolant, the effect of evacuating molten salt on the core is smaller and more beneficial to the engineering realization. Compared to one-time loading scheme, batch loading scheme can ensure that a small core excess reactivity in whole life-time which is easier to be controlled, but more complex to operation. In one-time loading scheme, the second shutdown system gains enough fast shutdown margins by increasing the number of second control rod, instead of reducing the stack height of core activity. The shutdown margins of both the first shutdown system and the second shutdown system meet the design requirements.

Responding to uncertain threat: A potential mediator for the effect of mindfulness on anxiety

Mindfulness-based interventions have gained extensive support for their application in the treatment of anxiety. However, their mechanisms remain largely unexplored. Excessive reactivity to uncertainty plays a central role in anxiety, and may represent a mechanism for the effect of mindfulness on anxiety, as mindfulness training fosters an open and accepting stance towards all aspects of experience. The present study sought to investigate both (i) self-reported intolerance of uncertainty (IU) as well as (ii) physiological and subjective responding to uncertain threat in a threat-of-shock paradigm, the NPU-threat test, as mediators for the relationship between mindfulness and anxiety in a cross-sectional study of healthy participants (N = 53). The results indicated that IU mediated the effect of mindfulness on some anxiety symptoms. In contrast, scores of physiological as well as subjective responses to uncertain threat from the NPU-threat test were largely unrelated to mindfulness, anxiety, or the IU self-report measure. The results provide initial evidence that reactions to uncertainty may play a role in the mindfulness-anxiety relationship and suggest that studies are needed to address how methodological variations of the NPU-threat test affect perceived levels of uncertainty and uncertainty-related anxiety.

Computational burnup analysis of the TRIGA Mark II research reactor fuel

In this study, analysis of the complete operational history of the “Jožef Stefan” Institute (JSI) TRIGA reactor was performed. Reactor power changes, core configurations and weekly excess reactivity measurements were analysed to obtain the needed data for fuel burnup calculations. More than 50 years of reactor operation was simulated using deterministic code TRIGLAV and stochastic code Serpent-2. The calculated core reactivities are in good agreement compared with the excess reactivity measurements. Code-to-code comparison is presented. Clear agreement is observed when comparing changes in core excess reactivity, and discrepancies are observed in the comparison of individual fuel element burnup and its isotopic composition. The Serpent-2 results are in better agreement with the measurements compared to the TRIGLAV code; nevertheless, a conclusion can be made that the TRIGLAV code is viable for TRIGA fuel management and burnup analysis. A three-dimensional (3D) burnup study was conducted, where individual fuel elements were further divided into multiple angular and axial depletion zones. Notable burnup effects were observed, and an explanation using surrounding water distance is presented.

Cardiac sympathetic activation and parasympathetic withdrawal during psychosocial stress exposure in 6-month-old infants.

Infant autonomic reactivity to stress is a potential predictor of later life health complications, but research has not sufficiently examined sympathetic activity, controlled for effects of physical activity and respiration, or studied associations among autonomic adjustments, cardiac activity, and affect in infants. We studied 278 infants during the repeated Still-Face Paradigm, a standardized stressor, while monitoring cardiac activity (ECG) and respiratory pattern (respiratory inductance plethysmography). Video ratings of physical activity and affect were also performed. Respiratory sinus arrhythmia (RSA) and T-wave amplitude (TWA) served as noninvasive indicators of cardiac parasympathetic and sympathetic activity, respectively. Responses were compared between infants who completed two still-face exposures and those who terminated after one exposure due to visible distress. Findings, controlled for physical activity, showed robust reductions in respiration-adjusted RSA and TWA, with more tonic attenuation of TWA. Infants completing only one still-face trial showed more pronounced autonomic changes and less recovery from stress. They also showed elevated minute ventilation, suggesting hyperventilation. Both reductions in adjusted RSA and TWA contributed equally to heart rate changes and were associated with higher negative and lower positive affect. These associations were more robust in the group of distressed infants unable to complete both still-face trials. Thus, cardiac sympathetic activation and parasympathetic withdrawal are part of the infant stress response, beyond associated physical activity and respiration changes. Their association with cardiac chronotropy and affect increases as infants' distress level increases. This excess reactivity to social stress should be examined as a predictor of future cardiovascular disease.

Исследование запаса реактивности в инновационных реакторах на быстрых нейтронах большой мощности в ЗЯТЦ

Исследование запаса реактивности в инновационных реакторах на быстрых нейтронах большой мощности в ЗЯТЦ

Neutronic performance of a thorium based mixed oxide fuel in a burner sodium-cooled fast reactor

The paper discusses the feasibility of using the Thorium Mixed-Oxide fuel with transuranic isotopes in a medium-sized Sodium-Cooled Fast Reactor core dedicated for actinide incineration with a capacity of 1000MWth. The core was designed based on the Advanced Burner Reactor 1000MWth studied in OECD/NEA SFR Benchmark. The SERPENT 2.1.31 Monte Carlo computer code with ENDF library was used to perform static criticality, isotopic depletion and safety parameters calculations for five consecutive fuel cycles. The key core parameters, eigenvalue, effective delayed neutron fraction, control rod worth, sodium void, doppler reactivity coefficients and isotopic inventories were investigated at the Beginning-of-Cycle (BOC) and End-of-Cycle (EOC). The results were compared with the medium MOX burner and large thorium core. Thanks to the internal U-233 breeding, the excess reactivity remained positive up to 700 days in comparison to 400 days for the reference core. The core has the ability to operate long campaigns, and at the same time, it is the efficient transuranic burner, with the conversion ratio ∼1. The penalty of the improved cycle was reduced Doppler effect and effective delayed neutron fraction. In comparison to the MOX reference core, sodium void effect was reduced.

Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel

The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

Desain Konseptual Reaktor Cepat Berpendingin Karbondioksida dan Berbahan Bakar Uranium Alam Dengan Daya 2400 MW

This paper presents the design concept of a carbon dioxide-cooled fast reactor. This reactor utilize U-10%Zr as fuel and SS316 as cladding. The strategy of modified CANDLE ( Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production ) was applied for burnup in the core with power 2400 MW. The reactor core calculations were performed with a cylindrical geometry that is varied on the height and diameter of the core using a set of CITATION and PIJ modules on the SRAC ( Standard Reactor Analysis Code ) program. The ideal core size was obtained with a high of 350 cm, and a diameter of 240 cm with the resulting survey parameter are effective multiplication factor( k eff ), excess reactivity, radial and axial power distribution, and power peaking. The reactor core reaches a critical condition with k eff 1.05 and excess reactivity 5.3% and radial power peaking 1.73. Optimization was done with power flattening, that is by dividing the core into two parts with a fuel fraction of 60% for the inner part with thick of 80 cm and fuel fraction of 65% for the outer part with thick of 40 cm, the results are 1.013, 1.3% and 1.5 for k eff , excess reactivity, and radial power peaking, respectively.

Novel homogeneous burnable poisons in pressurized water reactor ceramic fuel

Due to excess reactivity, fresh nuclear fuel often contains burnable poisons. This research looks at six different burnable poisons and their impacts on reactivity, material attractiveness, and waste management. An MCNP simulation of a PWR fuel pin was performed with a fuel burnup of 60 GWd/MTHM to determine when each burnable poison fuel type would decrease below a k∞ of 1. For determining the plutonium material attractiveness in each burnable poison fuel type, the plutonium isotopic content of the used fuel was evaluated using Bathke's Figure of Merit formula. For the waste management analysis, the thermal output of each burnable poison fuel type was determined through ORIGEN decay simulations at 100 and 300 years after being discharged from the core. The performance of all six burnable poisons varied over the three criteria considered and no single burnable poison performed best in all three considerations.

BANDUNG TRIGA 2000 REACTOR POWER ANALYSIS AS A FUNCTION OF THE NUMBER OF FUEL ELEMENTS AND THE POWER PEAKING FACTOR

The reactivity value of the Bandung TRIGA 2000 reactor core has decreased over time, so the power generated by the reactor is also getting smaller, despite the control rod position is fully withdrawn. Therefore, it is necessary to reshuffle and refuel the fuel element to increase the excess reactivity by considering the safety parameters, such as axial and radial power peaking factors, DNBR, dTsat, and temperature on the cladding and in the center of the fuel element. The analyzed reactor safety parameters are the number of fuel elements, which varied at 105, 110, and 115 elements, as well as power peaking factor, which varied at 1.55, 1.65, 1.75, 1.85, and 1.95. The calculations were done using MCNP and COOLOD-N2 programs. If DNBR ≈ 1.3 is determined as the safety limit for the operation of the Bandung TRIGA 2000 reactor, at PPF 1.95 (105, 110, and 115 fuel elements), it can be considered to operate the reactor at the power of 600-700 kW. However, at PPF of 1.75 (105, 110, and 115 fuel elements), the reactor can be operated at the power of 700-800 kW, and at PPF of 1.55 (105, 110, and 115 fuel elements), the reactor can be considered for operation at the power of 800-900 kW. The results of these calculations can be used for consideration in determining the operating limits of the Bandung TRIGA 2000 reactor.Keywords: TRIGA 2000, fuel element, power peaking factor, DNBR, boiling

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

KRUSTY Experiment: Reactivity Insertion Accident Analysis

The centerpiece of the Kilopower Project, i.e., the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) test, consists of the development and testing of a ground technology demonstration of a small fission power system based on a 1-kW(electric) space science power requirement. The KRUSTY test was authorized by the U.S. Department of Energy’s (DOE’s) National Nuclear Security Administration Nevada Field Office. Authorization was obtained by adding an amendment to the existing regulatory documents for the National Criticality Experiments Research Center to cover the KRUSTY experiment. This amendment was reviewed and approved by the DOE. The most important safety question for the experiment was the addition of over 2 $ of excess reactivity to the reactor system. This amount of excess reactivity meant that the analyst could postulate accidents where the reactor went prompt critical, leading to physical shock or melting of the fuel. This paper analyzes these accidents using computer calculations and examines the controls used to mitigate them. The estimation of the impacts both on accident progression and consequences of reactivity insertion events was a significant part of obtaining approval for the KRUSTY experiment. The regulatory approval of KRUSTY was one of the first to be obtained for a completely new reactor concept in many decades.

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Cold Critical Measurements

For the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) cold critical experiments, the KRUSTY component critical configuration was modified by the addition of parts that would be required for cold, warm, and hot critical experiments (including the vacuum chamber as well as the heat pipes and associated parts). Reactivity measurements were performed on the KRUSTY cold critical experimental configurations with the goal of obtaining reactivity-worth measurements on the beryllium oxide (BeO) reflector and the boron carbide (B4C) control rod parts. The resulting data are consistent and allow for accurate identification of the BeO and B4C part thicknesses required to achieve the excess reactivity needed for the KRUSTY warm and hot experimental configurations.

A novel protocol to measure startle magnitude in sheep

Temperament is a trait of interest as it impacts ease of handling, production and reactivity to stressful situations. Excessive behavioural reactivity, as an aspect of temperament, may lead to reduced welfare and stockperson safety. Current tests of behavioural reactivity involve isolating the subject and it can be difficult to differentiate between behavioural responses. We therefore aimed to develop a startle test in sheep which quantified variability in behavioural reactivity whilst addressing the effects of social isolation. The startle response is a consistent universal reaction to sudden and intense stimulation, however, magnitude of startle can be highly variable and may be modulated by affective state. In this study, 60 ewes were allocated into three social groups: isolated; surrounded by six life-sized pictures of sheep or; three conspecifics penned on two sides (total six). Sheep were tested over two days in sets of 10 per treatment. Sheep were confined to a 2.1 m × 1.2 m area, with a bowl of feed at one end, for four minutes (min). Approximately 15 s after the sheep commenced eating, a sudden one sec blast of compressed air was delivered to the face. Tri-axial accelerometers, visual scoring, agitometer scores and body temperature were used to assess the magnitude of the startle response and the pre- and post-startle fear behaviours. In the 15 s pre-startle, isolated sheep spent less time eating (p < 0.001; mean percentage of time: 37.4% vs 71.2% and 83.3%) and were 2–4 times longer in a vigilant state (p < 0.001; mean percentage of time: 38.9% vs 18.2% and 11.3%) than those with pictures and conspecifics respectively. Sheep in the conspecific treatment were 3–4 times faster to return to eating after startle than those in the isolated and picture treatments and there was a tendency for them to spend more time eating (p = 0.09) and less time vigilant (p = 0.09) post-startle. For each measure of startle magnitude (startle force, startle duration or retreat distance) the effect of treatment was not statistically important (p > 0.22), however each measure was positively correlated to time spent vigilant post-startle and there were significant increases in time to return to eating for every one unit increase in each of the startle magnitude measures independent of social treatment. The use of conspecifics did not provide social facilitation or buffering of the startle response. This shows promise as a behavioural reactivity test which can utilise conspecifics to reduce any potential negative effects of social isolation during testing.

Design Study of Gas Cooled Fast Reactor (GFR) with Uranium Plutonium Carbide (UC-PuC) as Fuel with Addition Protactinium (Pa-231)

Analysis performance of uranium plutonium carbide (UC-PuC) as fuel in gas cooled fast reactor (GFR) with addition of protactinium as a burnable poisons has been done. Neutronic analysis in this research was carried out using the SRAC code from JAERI with a nuclear library based on JENDL 4.0. The calculation is carried out by two steps, the first step is the PIJ calculation which calculates the fuel cell and the second step is the CITATION calculation which calculates the various configurations of the reactor core. The first calculation determines the k-eff value in a homogeneous core configuration. The results obtained show that the percentage of 10% is the sloping result with a k-eff value of 1%. The second calculation determines the k-eff value in the heterogeneous core configuration. The results obtained indicate that the fuel variation 8% -10% -12% is the most critical percentage with a peak power density value of less than 100 Watt/cc. Furthermore, the addition of protactinium with a variation of 0% to 5%. At a protactinium 4% percentage and 63% fuel fraction, the excess reactivity value is 1.02% or close to 1% which indicates that the reactor is in a critical condition.

Feasibility of an innovative long-life molten chloride-cooled reactor

Molten salt-cooled reactor is one of the six Gen-IV reactors with promising characteristics including safety, reliability, proliferation resistance, physical protection, economics, and sustainability. In this paper, a small innovative molten chloride-cooled fast reactor (MCCFR) with 30-year core and a target 120-MWt thermal power was presented. For its feasible study, neutronics, thermal-hydraulics, and radiation damage analysis were performed. The key design properties including kinetics parameters, reactivity swing, reactivity feedback coefficients, maximum accumulated displacement per atom (DPA) of reactor pressure vessel (RPV) and fuel cladding, and maximum coolant, cladding, and fuel temperatures were evaluated. The results showed the MCCFR could operate without refueling for 30 years with overall negative reactivity feedback coefficients up the end of its life. During its 30-year life, the excess reactivity was well managed by constantly pulling out the control rods. The maximum accumulated DPA on RPV and fuel cladding were 8.92 dpa and 197.03 dpa, respectively, which are both below the design limits. Similarly, the maximum coolant, cladding and fuel center temperatures were all below the design limits during its entire lifetime. According to these results, the MCCFR core design with long life is feasible.

Design Conceptual of 800MWt Long Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

A long-life pressurized water reactor (PWR) has been reviewed as an innovative reactor design that can fulfill electricity demand. This study aimed to find out the optimum design of 800MWt long life PWR using Thorium-Uranium dioxide (Th-U)O2 fuels with Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. An established computer code of SRAC 2006 with JENDL 4.0 as data nuclear library had been used for the analysis. A two-dimensional R-Z geometry and fuel volume fraction of 40% were used for core geometry analysis. The different fraction of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 had been carried out. The result of this study was a design of PWR 800MWt using Uranium dioxide fuel of 60% with enrichment 11%-12%-13% Uranium-235 and the addition of 0,025% Gd2O3 and 1,0% Pa-231 that could operate for ten years without refueling. The reactor could produce a power density of 45,4 watts/cc with excess reactivity about 3,6% dk/k. This study is expected to be a reference for a long-life pressurized water reactor using the Thorium-Uranium fuel cycles.