Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U–Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm³, 7.74 gU/cm³ and 8.57 gU/cm³. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Prospects of using different clad materials in a material test research reactor – Part 3 – The dynamic behavior

The effects of using different clad materials on the dynamics of a material test research reactor were studied. For this purpose, the aluminum clad of an MTR was replaced separately with stainless steel-316 and zircaloy-4. Simulations were carried out to determine the reactor performance under reactivity insertion and loss-of-flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that during the fast reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 59.34 MW, while stainless steel-316 cladded attained 48.74 MW and zircaloy-4 cladded attained maximum power of 55.87 MW. During the slow reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 12.38 MW, while stainless steel-316 cladded attained 12.23 MW and zircaloy-4 cladded attained maximum power of 12.34 MW. During the loss-of-flow transients, the reactor power of the stainless steel-316 cladded fuel remained slightly lower than the other two. The fuel temperature of stainless steel-316 and zircaloy-4 cladded fuels remained higher due to poor fuel–clad gap conductance.     

Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm³ by the fuel of density 4.00 gU/cm³ then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm³ in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29–62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1.     

Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling

Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650 °C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6 mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156 W/cm³ and its equivalent diameter is only 2.7 m for 1000 MWe class reactor core. High average core outlet temperature of 500 °C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost.     

Neutronic calculations of the U3Si LEU as a dispersed fuel for Low-Power Research Reactors

Usability of the LEU U₃Si dispersed fuel together with the actual UAl₄–Al HEU fuel (mixed core) in Low-Power Research Reactors (LPRRs) (～30 kW) was assessed in this paper. The use of both fuels together (33% HEU and 67% LEU) in LPRRs seems to be achievable from the neutronic point of view. High Initial Excess Reactivity (IER) can be achieved. To maintain the reactor performance in terms of neutron flux value in the internal and external irradiation sites the reactor power needs to be increased to about 32 kW. However the safety margin of the mixed core is smaller in both normal and accidental operation conditions.     

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl_x–Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si–Al and U₃Si₂–Al) and oxide (U₃O₈–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 °C to 50 °C and 100 °C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U₃O₈–Al was about 2% more than the original UAl_x–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.     

Measured and simulated reactivity insertion transients characteristics of NIRR-1

The power excursion characteristics of the Nigeria Research Reactor-1 (NIRR-1) under different reactivity insertion transients have been calculated using PARET/ANL 7.3 code. Experimental data of dynamic experiments performed in NIRR-1 facility during initial startup were used to benchmark the calculated data. For ramp insertion of total cold core excess reactivity of 3.77 mK, the reactor quickly reaches a maximum power of 82.3 kW in 390 s and decreases gradually due to the inherent safety features. The calculated maximum power for this insertion is in good agreement with a measured value of 79.4 kW recorded during the commissioning of NIRR-1. Similarly, moderator temperature for the hottest channel with respect to this transient was calculated to be 63.7 °C which is comparable to a measured value of 63.9 °C. Furthermore, step reactivity insertions of 2.23 mK and an approximately 0.32 mK, a regular criticality experimental check carried out at MNSR facilities by Safeguards Inspectors were accurately simulated.     

Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U–Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.     

Measurements of the isothermal,power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Breeding and void reactivity analysis on heavy metal closed-cycle water cooled thorium reactor

Design parameters of heavy water (D₂O) cooled thorium breeder reactors for actinides closed-cycle cases have been investigated to find a design feasible area of breeding and negative void reactivity. Heavy metals (HMs) closed-cycle shows narrower feasible area compared with feasible area of ²³³U closed-cycle. In thorium fuel cycle, the breeding capability of the reactors becomes worse when all HMs are recycled. The result shows an opposite profile of breeding capability compared with uranium fuel cycle which obtains higher breeding capability when more HMs are recycled. Feasible design area which has a breeding and negative void reactivity can be estimated for higher burnup, even higher than 60 GW d/t for ²³³U closed-cycle; however, it is limited to 36 GW d/t for HM closed-cycle. Contribution of capture ²³⁵U is more significant to reduce breeding capability and contribution of ²³⁴U is also more effective to make the reactor more positive or less negative void coefficient for HM closed-cycle case in thorium fuel cycle system.     

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO₂ and U–9Mo) with existing HEU fuel (UAl₄–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl₄–Al, UO₂ and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of ²³⁵U, respectively. Relative to 0.042 g ²³⁹Pu produced in case of UAl₄–Al HEU core, UO₂ and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO₂ based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.     

Neutronic analysis for core conversion (HEU–LEU) of Pakistan research reactor-2 (PARR-2)

Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl₄–Al). For core conversion, UO₂ fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO₂ of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

Radiation shielding mass saving for the magnet coils of the VISTA spacecraft

Radiation shielding structure of a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit has been investigated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle. The nuclear heating in the super conducting magnet coils determines the radiation shielding mass of the spacecraft. It was possible to achieve considerable mass saving with respect to a recent design work, coupled with higher design limits for coil heating (up to 5 mW/cm³). The neutron and γ-ray penetration into the coils is calculated using the S_N methods with a high angular resolution in (r–z) geometry in S₁₆P₃ approximation by dividing the solid space angle in 160 sectors. Total peak nuclear heat generation density in the coils is calculated as 3.143 mW/cm³ by a fusion power of 17 500 MW. Peak neutron heating density is 1.469 mW/cm³ and peak γ-ray heating density is 1.674 mW/cm³. However, volume averaged heat generation in the coils is much lower, namely 74, 163 and 337 μW/cm³ for neutron, γ-ray and total nuclear heating, respectively. The net mass of the radiation shielding for the magnet coils is 200 tonne by a total mass of 6000 tonne of the space craft.     

Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR

The sensitivity of various safety parameters, affecting the reactivity insertion limits imposed by clad melting temperature for a typical pool type research reactor, have been investigated in this work. The analysis was done for low enriched uranium (LEU) core with scram disabled conditions. The temperature coefficients of fuel and coolant, void/density coefficient and β_eff were individually varied and the reactor behavior for different ramp reactivity transients was studied. In this work ramp reactivity insertions from 1.6 to 2 $/0.5 s were selected and peak power, maximum fuel, clad and coolant temperatures were determined. Results show that peak power decreases with an increase in the Doppler coefficient of reactivity. However, it rises with an increase in the reactivity insertion. Core remains insensitive to the coolant temperature coefficient of reactivity for ramps in the range of 1.6–1.9/0.5 s. Peak power decreases with an increase in the void coefficient of reactivity (0.1 $/%void to 0.8 $/%void). With a decrease in the void coefficient of reactivity, the maximum fuel and clad temperatures show a non-linear rise. Power and temperature peaks in the transient are sensitive to the values of β_eff. Finally, it can be concluded that LEU is a safe core due to its smaller β_eff, larger Doppler coefficient and void coefficient of reactivity. It is inferred through this work that reactivity insertion limits of LEU core are quite insensitive to β_eff, the Doppler coefficient and the coolant temperature coefficient of reactivity. They are highly sensitive to the change of the void coefficient of reactivity in the core.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

Transmission electron microscopy characterization of irradiated U–7Mo/Al–2Si dispersion fuel

The plate-type dispersion fuels, with the atomized U(Mo) fuel particles dispersed in the Al or Al alloy matrix, are being developed for use in research and test reactors worldwide. It is found that the irradiation performance of a plate-type dispersion fuel depends on the radiation stability of the various phases in a fuel plate. Transmission electron microscopy was performed on a sample (peak fuel mid-plane temperature ～109 °C and fission density ～4.5 × 10²⁷ f m^?3) taken from an irradiated U–7Mo dispersion fuel plate with Al–2Si alloy matrix to investigate the role of Si addition in the matrix on the radiation stability of the phase(s) in the U–7Mo fuel/matrix interaction layer. A similar interaction layer that forms in irradiated U–7Mo dispersion fuels with pure Al matrix has been found to exhibit poor irradiation stability, likely as a result of poor fission gas retention. The interaction layer for both U–7Mo/Al–2Si and U–7Mo/Al fuels is observed to be amorphous. However, unlike the latter, the amorphous layer for the former was found to effectively retain fission gases in areas with high Si concentration. When the Si concentration becomes relatively low, the fission gas bubbles agglomerate into fewer large pores. Within the U–7Mo fuel particles, a bubble superlattice ordered as fcc structure and oriented parallel to the bcc metal lattice was observed where the average bubble size and the superlattice constant are 3.5 nm and 11.5 nm, respectively. The estimated fission gas inventory in the bubble superlattice correlates well with the fission density in the fuel.     

Test irradiations of full-sized U3Si2-Al fuel plates up to very high fission densities

In the course of the licensing procedure of the ‘Forschungsneutronenquelle Heinz Maier-Leibnitz’, i.e. the new 20 MW high-flux research reactor FRM II in Garching near Munich, extensive test irradiations have been performed to qualify the U₃Si₂-Al dispersion fuel with a relatively high density of highly enriched uranium (93 wt% of ²³⁵U) up to very high fission densities. Two of the three FRM II type fuel plates used in the irradiation tests contained U₃Si₂-Al dispersion fuel with HEU densities of 3.0 gU/cm³ or 1.5 gU/cm³ (‘homogeneous plates’) and one plate had two adjacent zones of either density (‘mixed plate’). They were irradiated in the French MTR reactors SILOE and OSIRIS in the years before 2002. The local plate thickness was measured on certain tracks along the plates during interruptions of the irradiation. The maximum fission density obtained in the U₃Si₂ fuel particles was 1.4 × 10²² f/cm³ and 1.1 × 10²² f/cm³ in the 1.5 gU/cm³ and 3.0 gU/cm³ fuel zones, respectively. In the course of the irradiations, the plate thickness increased monotonically and approximately linearly, leading to a maximum plate thickness swelling of 14% and 21% and a corresponding volume increase of the fuel particles of 106% and 81%, respectively. Our results are discussed and compared with the data from the literature.     

Thermal,structural, and radiological properties of irradiated graphite from the ASTRA research reactor – Implications for disposal

The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry/synchrotron powder X-ray diffraction between 25 °C and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from ∼10¹⁷ to ∼10²⁰ n cm⁻² over the life time of the reactor at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 25 to 572 J g⁻¹ and a Wigner energy accumulation rate of ∼7 × 10⁻¹⁷ J g⁻¹/n cm⁻². The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 °C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence >1.5−5.0 × 10¹⁹ n cm⁻²) is destroyed while that of samples farther from the reactor core (fast-neutron fluence <1.5−5.0 × 10¹⁹ n cm⁻²) is intact, with marked swelling along the c-axis. The dependence of the c lattice parameter on temperature between 25 °C and 200 °C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c-axis of ∼ 26 × 10⁻⁶ °C⁻¹. However, at 200 °C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The ¹⁴C activity in the inner thermal column graphite ranges from 6 to 467 kBq g⁻¹. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 °C for 24 h in a hot-cell facility prior to interim storage.