Obtaining molybdenum-99 in the IRT-T research reactor using resonance neutrons

The yield of ⁹⁹Mo from the ⁹⁸Mo(n,γ)⁹⁹Mo reaction significantly depends of the energy spectrum of the neutron flux. It is well known that the cross-section for this reaction is about 130 mb, whereas the resonance integral of the reaction is 6.9 b. The aim of this work was to investigate the conditions that let to increase ⁹⁹Mo yield from the targets with natural and enriched isotope composition under irradiation by resonance neutrons at the IRT-T research reactor.The calculations of integrated cross-sections of all Mo isotopes in the region of the ⁹⁸Mo resonances showed that screening in the target with natural isotope composition by other isotopes is relatively small. So the ⁹⁸Mo in the natural mixture can be activated by resonance neutrons approximately in the same manner as pure ⁹⁸Mo.Experimental measurements of the ⁹⁸Mo(n,γ) effective cross-section using the MoO₃ sample with natural and enriched composition in the reactor channels with the beryllium moderator with the thickness of 20 up to 90 mm showed that the effective cross-sections in these channels reach the value of 700 mb. The contribution of the epithermal neutrons into the ⁹⁸Mo activity was 68% for the enriched targets and 78% for natural molybdenum, respectively.At that channel it is possible to produce ⁹⁹Mo with specific activity up to 3.4 Cu/g with samples of natural isotope composition and up to 15 Cu/g with enriched samples on the base of reactors with neutron flux of (1.7 × 10¹⁴ n/(cm² s)). Such ⁹⁹Mo specific activity is enough not only to realize extraction technologies production of ^99mTc, but to manufacture sorption generators of ^99mTc without wastes.     

Thermodynamic considerations and prediction of the primary coolant activity of Tc

A physical model has been developed to describe the coolant activity behaviour of ⁹⁹Tc, during constant and reactor shutdown operations. This analysis accounts for the fission production of technetium and molybdenum, in which their chemical form and volatility is determined by a thermodynamic treatment using Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix and vaporization from the fuel-grain surface. Based on several in-reactor tests with defective fuel elements, and as supported by the thermodynamic analysis, the model accounts for the washout of molybdenum from the defective fuel on reactor shutdown. The model also considers the recoil release of both ⁹⁹Mo and ⁹⁹Tc from uranium contamination, as well as a corrosion source due to activation of ⁹⁸Mo. The model has provided an estimate of the activity ratio ⁹⁹Tc/¹³⁷Cs in the ion-exchange columns of the Darlington Nuclear Generating Station, i.e., 6 × 10⁻⁶ (following ∼200 days of steady reactor operation) and 4 × 10⁻⁶ (with reactor shutdown). These results are consistent with that measured by the Battelle Pacific Northwest Laboratories with a mixed-bed resin-sampling device installed in a number of Pressurized Water Reactor and Boiling Water Reactor plants.     

New method of producing <Superscript>99</Superscript>Mo in molten-salt fluoride fuel

The status and the problems of world ⁹⁹Mo production are presented. A comparative analysis is made of reactor methods of ⁹⁹Mo production. It is noted that the currently used technologies and research reactors are not satisfying the growing demand in medicine for this isotope. It is underscored that the role of alternative production technologies has grown. In the development of new ⁹⁹Mo production technologies, the experimental results obtained on the basis of research program conducted on the MSRE reactor with molten-salt fluoride fuel have been analyzed. The analysis revealed a special behavior of certain fission products including ⁹⁹Mo: they leave the melt spontaneously and enter the gas phase. The authors hypothesize that highly volatile fluorides of the indicated products are formed in the melt; this explains the effect indicated. The effect is used as a basis to propose a new reactor method of producing fissionproduced ⁹⁹Mo. Concrete examples of a way to implement the new method of producing fission-produced ⁹⁹Mo using molten-salt fluoride nuclear fuel are presented.     

Radiochemical Determination of Fission Rate in JRR-4

The fission rate in the core of the Japan Research Reactor 4 (JRR-4) was determined by a method based on radiochemical analysis of ⁹⁹Mo formed in the U samples irradiated in the reactor core.

The contribution of epithermal neutron fission to the total fission rate was evaluated from the Cd ratio for U fission. The contribution was several percent.

For comparison, the thermal neutron flux also was measured, by Au-foil activation. The fission rate determined from the U samples agreed well with the Au-foil data, except at positions in the peripheral region of the reactor core.     

PIK-2 Reactor with a Low Consumption of High-Enrichment Uranium

Calculations of the possibility of switching the matrix of the fuel-element cores and the vessel with the reactor envelope to weakly neutron-absorbing aluminum and lowering at the same time the fuel content in the fuel elements by 40% are performed for the high-flux PIK reactor which is under construction in Gatchina. As a result, the flux density of the thermal neutrons in the heavy-water reflector increases by a factor of 1.4–1.5 and the maximum unperturbed flux reaches a world record value – 1.9·10¹⁵ sec^–1·cm^–2. The nonuniformity of the volume energy-release decreases to 2.3. The weight of the control rods and reactor safety increase. As a result of the decrease of the fuel content in the fuel elements and the increase of the run duration, about 60 kg of high-enrichment uranium ²³⁵U are saved annually, decreasing annual fuel costs by one third.     

Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enriched uranium foil (<20% ²³⁵U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for the production of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at PINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data shows that annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.     

A comparative neutronic study of the standard HEU core and various potential LEU alternatives for a typical MNSR system

A comparative study has been performed for neutronic analysis of highly enriched in uranium (HEU) and potential low enriched in uranium (LEU) cores for the Pakistan Research Reactor-2 (PARR-2) taken as a typical miniature neutron source reactor (MNSR) system. The group constant generation has been carried out using transport theory code WIMS-D4 and a detailed five-group RZ-model has been used in the CITATION code for multigroup diffusion theory analysis. The neutronic analysis of the 90% HEU reference and potential LEU alternative: UO₂, U₃Si₂ and U9Mo, cores has been carried out yielding 11%, 20.7% and 14.25% enrichments with corresponding values of excess reactivity: 4.33, 4.30 and 4.07 mk. These results have been found in good agreement with recently reported Monte Carlo-based transport theory calculations. The diffusion theory-based calculated values of thermal flux profiles for axial as well as for radial directions have been found to agree well with the corresponding experimental measurements. The UO₂-based LEU core has been found having flux spectrum closest to the reference core while U9Mo core has significantly harder flux spectrum at irradiation site.     

A pulsing reactor

A proposal for the design of a pulsing reactor operating on light water and enriched uranium is given in this article, and the method of calculating the physical and thermal parameters of such a reactor are indicated. It is shown that when certain definite conditions are established very stable heat generation during the neutron pulses may be obtained. In the version of a pulsating reactor discussed here, the calculations indicate that it is possible to obtain over 5 Mw power with an average thermal neutron flux greater than 10¹⁴ neutrons/cm²· sec and a maximum neutron flux (during neutron pulse) of approximately 10¹⁷ neutrons/cm² · sec in the reactor core.The author takes this opportunity to express gratitude to Academician I. V. Kurchatov for his interest in this work. The author is also grateful to U. N. Zankov for his discussion of the conclusions, and to A. K. Sokolov for his participation in a number of calculations.     

Fission MOLY production at PARR-1 using LEU plate type target

Neutronic analyses and depletion calculations have been performed for the production of ⁹⁹Mo at PARR-1. Analysis has been performed with 20% enriched ²³⁵U target bearing plate type aluminized fuel (U-Al). Target (target holder and fuel plates) design contains three fuel plates and two aluminum dummy plates. Neutronic calculations were carried out for core at the beginning of equilibrium cycle of Pakistan Research Reactor-1 (PARR-1). Target analysis was performed by irradiating it at a location of maximum thermal flux available in the core. For this purpose, irradiation was performed at five different axial planes of the central water box facility. Thermal neutron flux profiles were also studied at different axial positions in available irradiation locations. Computer code WIMSD/4, a transport theory lattice code was employed for the generation of 10 group microscopic cross-sections. Diffusion theory code CITATION was utilized for three-dimensional modeling of the core. It was observed that from reactivity point of view, insertion or removal of target from the core will not affect the safety of reactor. Maximum heat flux in the target would be 102.68 W/cm² which is below the point of onset of nucleate boiling. However, forced flow is required to avoid initiation of nucleate boiling. The computer code ORIGEN2 was employed for depletion calculations. Analysis was performed for 100 Ci activity of ⁹⁹Mo. After 100 days decay, waste activity will be less than 1 Ci and it will not pose any problem for handling radioactive waste.     

Post-irradiation examination of uranium-7 wt% molybdenum atomized dispersion fuel

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm² while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% ²³⁵U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al₃ and (U,Mo)Al₄. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l’énergie atomique (CEA).     

Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.     

Research reactor RBT-6

Conclusions Reactor RBT-6 is simple in construction and is easily accessible for conducting experiments. The values of neutron flux in it are high for small thermal power; this together with the large duration of continuous operation ensures the possibility of conducting a wide range of experimental investigationa. Such a reactor may be recommended as a research reactor for irradiation of samples of materials up to moderate flux values (10¹⁹–10²¹ neutrons/cm²) and for conducting experiments for studying the change in the properties of materials during irradiation, which are becoming increasingly more important. Estimates show that the number of used fuel assemblies of the SM-2 reactor are sufficient for the operation of several such reactors.If necessary, the number of experimental channels in the active zone of such a reactor can be increased by increasing the number of fuel assemblies and the thermal power. Beryllium can be used as the lateral reflector. This results in a decrease of the volume of the active zone and the thermal power of the reactor, but increases its cost.Translated from Atomnaya Énergiya, Vol. 43, No. 1, pp. 3–7, July, 1977.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Thermal neutron cross-section and resonance integral of the Mo(n,γ)Mo reaction

We measured the thermal neutron cross-section and the resonance integral of the ⁹⁸Mo(n,γ)⁹⁹ Mo reaction by the activation method using a ¹⁹⁷Au(n,γ)¹⁹⁸ Au monitor reaction as a single comparator. The high-purity natural Mo and Au metallic foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The necessary correction factors for the γ-ray attenuation (F_g), the thermal neutron self-shielding (G_th) and the resonance neutron self-shielding (G_epi) effects, and the epithermal neutron spectrum shape factor (α) were taken into account. In addition, for the ⁹⁹Mo activity measurements, the correction for true coincidence summing effects was also taken into account. The thermal neutron cross-section for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction has been determined to be 0.136 ± 0.007 barn, relative to the reference value of 98.65 ± 0.09 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸ Au reaction. The present result is, in general, in good agreement with most of the experimental data and the recently evaluated values of ENDF/B-VII.0, JENDL-3.3, and JEF-2.2 by 5.1% (1σ). By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction is 7.02 ± 0.62 barn, which is determined relative to the reference values of 1550 ± 28 barn for the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction. The present resonance integral value is in general good agreement with the previously reported data by 8.8% (1σ).     

Reactivity analysis of solution reactors for medical-radioisotope production

One of the primary methods to produce medical isotopes, such as ⁹⁹Mo, is by irradiation of uranium targets in heterogeneous reactors. Solution reactors present a potential alternative to produce medical isotopes. The medical isotope production reactor concept has been proposed to produce medical isotopes with lower uranium consumption and waste than the corresponding fuel consumption and waste in heterogeneous reactors. Commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, fuel-solution temperature increase and radiolytic-gas bubble formation introduce a negative reactivity feedback that has to be mitigated. This work analyzes the reactivity effects on the operation of solution reactors for the production of medical isotopes and provides some reactor characteristics that may mitigate the negative reactivity feedback introduced by the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles.     

Neutron-Physical Calculation of a Fresh Zone in the Natural Nuclear Reactor at Oklo

A full-scale computer model of zone 2 of the Oklo reactor with a realistic materials composition is constructed. Modern Monte Carlo programs are used to calculate the multiplication factor, the excess reactivity, and the neutron flux for the fresh reactor zone 2. The calculations are performed in a wide range of variation of the zone parameters: the uranium content in the contemporary Oklo zone is varied from 35 to 55 mass% and the water content from 0.355 to 455 g/cm³. The power effect is determined. The temperature of the fresh zone is found to be 725 ± 55 K, at which the reactor is critical and can operate in a stable manner for a long time. The power of the reactor is maintained by negative feedback. The neutron spectrum, over which the cross section of strong absorbers must be averaged, for example, ₆₂ ¹⁴⁹ Sm, in order to obtain the most accurate bounds on the possible shift of the samarium resonance as a result of a change in the fundamental constants, is found for 700 K.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 306–316, April, 2005.     

Economy of uranium resources in a three-component reactor fleet with mixed thorium/uranium fuel cycles

The potential for minimizing uranium consumption by using a reactor fleet with three different components and mixed thorium/uranium cycles has been investigated with a view to making nuclear power a more sustainable and cleaner means of generating energy. Mass flows of fissile material have been calculated from burnup simulations at the core-equivalent assembly level for each of the three components of the proposed reactor fleet: plutonium extracted from the spent fuel of a standard pressurized water reactor (first component) is converted to ²³³U in an advanced boiling water reactor (second component) to feed a deficit of multi-recycled ²³³U needed for the Th/²³³U fuel of the light/heavy water reactor (third component) which has a high breeding ratio. Although the proposed fleet cannot breed its own fuel, we show that it offers the possibility for substantial economy of uranium resources without the need to resort to innovative (and costly) reactor designs. A very high fleet breeding ratio is achieved by using only currently existing water-based reactor technology and we show that such three-component systems will become economically competitive if the uranium price becomes sufficiently high (>300 $/kg). Another major advantage of such systems is a corresponding substantial decrease in production of plutonium and minor actinide waste.     

Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

Low enriched uranium foil (19.99% ²³⁵U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.     

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.