Target Burnup Research Articles

Heavy metal loading effects on special nuclear material and waste of an uprated Indonesian Reaktor Daya Eksperimental

This technical policy study investigated the effect of heavy metal loading (HML) quantity in the pebble fuel on the special nuclear material (SNM) and waste quantities in an uprated Indonesian pebble bed reactor design, the Reaktor Daya Eksperimental (RDE). A set of Monte Carlo simulations, performed using OpenMC, was deployed to simulate the pebble fuel depletion using an infinite lattice reactor calculation. This study considers HML, fuel residence time, leftover 235U, total Pu, medium- and long-lived wastes, number of depleted pebbles, and volume of waste at different HML values, at attainable discharged fuel burnup, at a target burnup of 80 GWd/MTU and at different reactor powers (10 and 40 MWt). Due to an over-moderation effect, a higher HML quantity shortened the attainable discharged fuel burnup level, which also translated to a lower fuel utilization. An 80% higher HML quantity resulted in about a 25% lower burnup level. Assuming a five-pass refueling scheme, this resulted in approximately three times more leftover 235U, 2.4 times more Pu, 1.4 times more medium-lived waste, and 1.5 times more long-lived waste collections per year. The study showed that a power uprating by a factor of four increased the SNM and waste quantities by four times.

PCM-based uncertainty reduction in support of model validation for depletion calculations

This manuscript further develops a recent methodology, denoted by Physics-guided Coverage Mapping (PCM), to support model validation for neutronic depletion calculations. The overarching goal of model validation is to develop confidence in model predictions for the application of interest via fusion of both simulation results and measurements from scaled-down experiments, and whenever possible to improve predictions by explaining the observed discrepancies. This manuscript focuses on the isotopic depletion problem, that is how to improve the predictions of depleted fuel isotopic across the range of expected burnup based on a limited number of post-irradiation measurements. PCM employs an information theoretic approach, capable of directly transferring, i.e., without performing model inversion, biases and their uncertainties from the available measurements to the quantities of interest (QoIs), representing the isotopic concentrations at different burnup values and/or different irradiation spectra. It precludes the need for sensitivity coefficients and only requires forward model executions, and can be applied using non-informative priors, often required by Bayesian-based methods. This is achieved via a mapping kernel relating a number of predictor variables, the concentrations of single or multiple isotopes at certain burnup, to the QoIs, the isotopic concentrations at target burnup such as end of life. Proof-of-principle calculations are demonstrated using both representative PWR and BWR lattice models, where the goal is to employ measurements at given burnup value(s) from one lattice to predict the isotopic concentrations across burnup for the same or the other lattice. Results show 50% to 90% reduction in uncertainties of isotopic concentrations across burnup as compared to the prior uncertainty.

Irradiation behaviour of alloy D9 clad and wrapper in FBTR

Titanium modified austenitic stainless steel 316 (Alloy D9) is the material of clad and wrapper for the 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam, India. Towards evaluating the irradiation performance of the mixed oxide (MOX) fuel design and Alloy D9 structural material, a test fuel subassembly (FSA) made of Alloy D9 wrapper with 37 MOX fuel pins of Alloy D9 cladding was irradiated in Fast Breeder Test Reactor (FBTR) to a peak burnup of 112 GWd/t. The dimensional changes, swelling and mechanical properties of alloy D9 clad and wrapper at displacement damage levels of 40-62 dpa were evaluated through post-irradiation examinations (PIE) and compared with that of AISI 316 clad/wrapper of FBTR. This study has revealed superior performance of alloy D9 compared to AISI 316 counterparts and suggests that MOX fuel subassembly of PFBR with Alloy D9 cladding and wrapper can safely attain the initial target burnup of 100 GWd/t and corresponding displacement damage of 85 dpa.

Burnup computations of online-refueling process for pebble-bed reactors using layer-mixed-shell fuel movement model

This study aims to propose a model for dynamically simulating the online-refueling process in pebble-bed reactor (PBR) using MCNPX. PBR has a special feature of online-refueling which can greatly reduce the outage time and enable a higher burnup in spent fuel. However, this feature also results in the dynamical fuel movements which may significantly increase the difficulty and the computational time in computer simulation. Therefore, an appropriate model is necessary to be proposed to simulate the burnup characteristics of online-refueling and to reduce the computational time simultaneously. All the calculations in this study were performed using MCNPX 2.7.0 with the ENDF/B-VII continuous energy nuclear data library. The PBR model was built according to the core design of HTR-10 but adopted some reasonable assumptions. The refueling process was emulated by utilizing the fuel loading scenario of the once through then out (OTTO) in combination with the layer-mixed-shell fuel movement. Additionally, the layer-mixed-shell fuel movement considered the concept of fuel channels, where the fuel pebbles only move in the same fuel channel, such that the burnup characteristics of fuel pebbles in both radial and axial direction can be identified separately. The core was divided into 9 fuel zones with a fixed volume and 3 fuel channels with a variety of fuel zones. Furthermore, the number of fuel zones in each fuel channel was determined based on the relative residence time of fuel pebbles in the core. The results revealed that the core can achieve an equilibrium fuel cycle after refueling several times, and after that all the core characteristics can nearly maintain unchanged between different cycles. Although the refueling process was modeled based on the OTTO fuel loading scenario instead of the multi-pass one, the discharged burnup can still reach the target burnup of the spent fuel for HTR-10, i.e. 72 GWd/tHM. In addition, the average discharged burnup under the equilibrium fuel cycle is approximate to 80 GWd/tHM, which also coincides the design value of the spent fuel for HTR-10. Therefore, the layer-mixed-shell movement model can consider the fuel movements in either radial or axial directions simultaneously and enable a more accurate prediction to the real refueling process of HTR-10 than our previous studies.

Evaluation on fuel cycle and loading scheme of the Indonesian experimental power reactor (RDE) design

Abstract In 2014, Indonesian government launched a program to build an Experimental Power Reactor, named Reaktor Daya Eksperimental (RDE), in South Tangerang, Banten. The RDE is a 10 MWt pebble-bed reactor utilizing TRISO fuel technology. A comprehensive neutronics and burnup performance evaluation on the RDE fuel loading schemes (once-through-then-out (OTTO) and multipass), fuel cycles (uranium and thorium) and optimum fuel compositions is conducted and reported in this paper. The optimum design parameters of the RDE are expected to contribute in minimizing the fuel cost, storage capacities, and other back-end environmental burden. The evaluation results show that OTTO fuel loading scheme is a better trade-off option considering the complexity of the mechanisms involved in the design. The optimal composition for uranium fuel cycle is 8 g heavy metal/pebble with U-235 content of 13.7 w/o for achieving the design target burnup of 80 GWd/t, while for thorium fuel cycle is 18.2 g/pebble with U-233 content of 7.2 w/o.

Spectral composition of thermonuclear particle and recoil nuclear emissions from laser fusion targets intended for modern ignition experiments

Determination of the spectra and yields of thermonuclear particles is an actual research line for inertial confinement fusion (ICF) in two areas: the development of diagnostic methods for ICF plasma and thermonuclear reactor design for energy production. The latter is particularly important for thermonuclear neutron emission, which contains the bulk of the evolving energy. This work is devoted to the determination and analysis of the energy spectrum composition of the emissions of thermonuclear neutrons, charged thermonuclear particles and recoil nuclei from ICF targets, which correspond to a laser pulse energy of ≈2 MJ and are designed for modern experiments to achieve a positive energy yield. The spectra are determined on the basis of hydrodynamic numerical simulations of ICF target burnup, modeling the generation of thermonuclear particles and their interaction with the target by means of the Monte Carlo method. The spectrum characteristics are discussed with reference to the problems of corpuscular diagnostics and radiation damage to the materials of thermonuclear reactor units.

Evaluation of Fuel-Clad Chemical Interaction in PFBR MOX test fuel pins

Fuel Clad Chemical Interaction (FCCI) is one of the life limiting issues in the MOX fuel pins of fast breeder reactors. Clad wastage due to FCCI coupled with stress arising from fission gas pressure and Fuel Clad Mechanical Interaction (FCMI) due to fuel swelling can lead to fuel pin failure. Non-destructive evaluation (NDE) techniques such as Gamma Scanning, Eddy current testing and Neutron Radiography have been successfully used on MOX fuel pins of a test sub-assembly irradiated in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GW d/t to detect signatures of FCCI. The results obtained by the various NDE techniques have been correlated and verified through metallographic examination on fuel pin cross-sections. The measured clad wastage of 85 μm agrees well with a model developed for MOX fuel with high Pu content. The results of examinations have enabled validation of the model and given confidence to the designer that PFBR MOX fuel can safely attain the target burn-up of 100 GW d/t.

A comparative study between flattening filter-free beams and flattening filter beams in radiotherapy treatment

Abstract Flattening filter-free (FFF) beams generated by medical linear particle accelerators (linacs) have recently been used in radiotherapy clinical practice. FFF beams have fundamental physical parameter differences with respect to standard flattening filter (FF) beams, such that the generally used dosimetric parameters and definitions are not always viable. This study investigates dosimetric parameters for use in the quality assurance of FFF beams generated by medical linacs in radiotherapy. The main characteristics of the photon beams are analyzed using specific data generated by a Varian TrueBeam linac having both FFF and FF beams of 6 and 10 MV (megavolt) energy, respectively. Definitions for dose profile parameters are suggested, starting from the renormalization of the FFF with respect to the corresponding FF beam. From this point, the flatness concept is translated into one of “un-flatness”, and other definitions are proposed, maintaining a strict parallelism between FFF and FF parameter concepts. The quality controls used in establishing a quality assurance program when introducing FFF beams into the clinical environment are given, maintaining similarity to those used for standard FF beams, and recommendations for the introduction of FFF beams into clinical radiotherapy application for breast cancer patients are provided as an example for comparison between FFF and FF for dose distribution and coverage for a target volume. Although there are many advantages of using a FFF beam, especially for advanced radiotherapy techniques, there are a few limitations (e.g., using a relatively higher energy photon beam for stereotactic radiotherapy (SRT), limited speed of current multileaf collimators (MLCs), and off-axis distance-dependent modulation in intensity-modulated radiation therapy (IMRT)) as well as challenges (e.g., criteria for beam quality evaluation and penumbra, establishment of dosimetry methods, and consequences of photon target burn-up) that need to be addressed for establishing the FFF beam as a viable alternative to the FF beam.

Evaluation of different physical parameters based on standard photon beam versus flattening filterfreein treatment cancer patients

cccc In radiotheraby, the purpose of using standard photon beam (Flattening Filter; FF) is to convert the forward peaked MV bremsstrahlung photon intensity into uniform intensity pattern for obtaining clinically acceptable beam profile. Recently a number of studies were carried out on existing medical linac by removing the flattening filter to produce the unflat photon beam and demonstrated their feasibility in the implementation of advanced radiotherapy techniques. Flattening Filter Free (FFF) beam have a fundamental physical parameter differences with respect to the standard filter flattened beams, making the generally used dosimetric parameters and definitions not always viable. In this concern, the current paper is a trial to shed further light on studying some dosimetric parameters for use in quality assurance of FFF beams generated by medical linacs in radiotherapy. The main characteristics of the photon beams have been analyzed using specific data generated by a Varian TrueBeamlinac having bothFFF and FF beams of 6 and 10 MV energy. Definitions for dose profile parameters are suggested starting from the renormalization of the FFF with respect to the corresponding FF beam. From this point the flatness concept has been translated into one of “unflatness” and other definitions have been proposed, maintaining a strict parallelism between FFF and FF parameter concepts. In summary, although there are a number of advantages of using a FFF beam especially for advanced radiotherapy techniques there are a few challenges (e.g., criteria for beam quality evaluation and penumbra, establishment of dosimetry methods, and consequences of photon target burn-up) which need to be addressed for establishing this beam as an alternate to the FF beam.

Design concept for a small pebble bed reactor with ROX fuel

The conceptual design of a small rock-like oxide fuel pebble bed reactor with once-though-then-out (OTTO) cycle is proposed here. TRISO-coated particles based on AGR-1 design were used to achieve a target burnup larger than 100GWd/t-HM without any failure of spent fuel. In the first step, optimization of fuel composition was implemented by cell calculations. After that, whole core calculations were performed with and without movement of the fuel pebbles. With a heavy metal amount of 2g per pebble and 20% uranium enrichment, the pebble bed reactor with OTTO cycle could achieve maximum burnup of about 145GWd/t-HM and fissions per initial fissile atom (FIFA) of 75%. The results show that the core height can be reduced due to the fact that the impact of bottom core on burnup performance is insignificant. Also, the peak power density of the reactor exceeded the limit of that for the PBMR design. Therefore, subsequent optimizations of the core design were carried out by decreasing the core height and reactor power to reduce the construction cost as well as the peak power density. A reactor with 6-m core height and 120-MWth reactor power was ultimately determined as the optimal design for a pebble bed reactor with ROX fuel. This optimal design also has a negative temperature coefficient, and the peak power density was less than the limit of 10W/cm3.

Reactivity Swing as a Function of Burnup for Uranium-Fueled Fast Reactors

In this study, the characteristics of changes in reactivity due to increasing burnup of uranium-fueled fast reactors are analyzed. A new classification system for nuclear reactor cores based on their uncontrolled tendency for reactivity changes during burnup was introduced and the design-optimization strategy for any fast reactor core aimed at a minimized reactivity swing is outlined. The 235U feed-fuel enrichment level that minimizes the burnup reactivity swing of a sodium-cooled metallic-fueled core is 10% to 12.5% for an average target fuel burnup of 1% to 20% FIMA (fission of initial metal atom). The higher the target burnup of the system, the lower the feed-fuel enrichment level that minimizes swing. The minimum attainable swing for a 125-MW(thermal) metallic-fueled sodium-cooled core is found to be ∼200 pcm for 5% FIMA burnup and increases to ∼800 pcm for a system aiming at 10% FIMA. In general, if the target discharge burnup is doubled, the minimum attainable burnup reactivity swing quadruples. Any optimized minimum reactivity swing core will form a positive parabolic uncontrolled reactivity trajectory with burnup, where the beginning of cycle and end of cycle reactivities are equal. Uranium-fueled fast cores with minimized burnup reactivity swing are net consumers of fissile material, with a fissile conversion ratio in the range of 0.7 to 0.9.

Neutronic feasibility study of U–Th–Pa based high burnup fuel for pebble bed reactors

Neutronic feasibility study of U–Th–Pa based high burnup fuel for the pebble bed reactors has been carried out. An additional content of P231a in a mixed U233–T232h fuel pebble was optimized for simultaneously controlling long-term excess reactivity and attaining high fuel burnup. Because of a large thermal absorption cross section, P231a has a potential to absorb thermal neutrons and act as a burnable absorber in the early stage of burnup. In the latter stage, P231a is transmuted to U233 and act as a fertile fuel. Parametric study was performed for heavy metal (HM) amount, U233 enrichment and P231a content. The objective is to attain a target burnup greater than 100 GWd/T with a low uranium enrichment and a small content of P231a. Six fuel compositions with the HM amount of 9–18 g per pebble, the U233 enrichment of 8 and 10%, and the P231a content of 3.2–5.5% were selected for discussion and comparison. The burnup levels of about 130–170 GWd/t are achieved for the six fuel compositions. The results of temperature coefficient show that negative reactivity feedback during burnup with the increase of temperature is assured, which is one of the important safety parameters.

On the practical aspects of large-scale production of 177Lu for peptide receptor radionuclide therapy using direct neutron activation of 176Lu in a medium flux research reactor: the Indian experience

This paper accentuates on the practical aspects and intricate technicalities involved in the large-scale production of 177Lu with specific activity >740 GBq mg−1 following (n,γ)177Lu route in a medium flux (~1.2 × 1014 n cm−2 s−1 thermal neutron) research reactor. The implication of target burn-up on the specific activity of 177Lu during irradiation was discussed in detail. 177Lu obtained from this route has been extensively utilized for targeted therapy in patients with neuroendocrine tumor in India. The important details available from our experience, as well as technical know-how, would be of considerable value for institutions planning to pursue 177Lu production through (n,γ)177Lu route.

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

The 500MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

An approach for evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems

One of the important issues in the study of Innovative Nuclear Energy Systems is evaluating the integrity of fuel applied in Innovative Nuclear Energy Systems. An approach for evaluating the integrity of the fuel is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes determining fuel life time were reviewed and fuel integrity was analyzed and compared with the failure criteria.Metal and nitride fuels with austenitic and ferritic stainless steel (SS) cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for Innovative Nuclear Energy Systems, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic–martensitic steel (PNC–FMS) and oxide dispersion strengthened steel (PNC–ODS).The analysis showed that the fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In case of ferritic steel, on the other hand, the fuel lifetime is controlled by cladding creep rupture. The lifetime evaluated here is limited to 200GWd/t, which is lower than the target burnup value of 500GWd/t. One of the possible measures to extend the lifetime may be reducing the fuel smeared density and ventilating fission gas in the plenum for metal fuel and by reducing the maximum cladding temperature from 650 to 600°C for both metal and nitride fuel.

Development of IFAC-1 SS: An Advanced Austenitic Stainless Steel for Cladding and Wrapper Tube Applications in Sodium-Cooled Fast Reactors

Fuel cycle cost of sodium cooled fast reactors (SFRs) is strongly dependent on the in-reactor performance of core structural materials, i.e., cladding and wrapper tube materials of the fuel subassembly, which are subjected to intense neutron irradiation during service, leading to unique materials problems like void swelling, irradiation creep and helium embrittlement. In order to increase the burnup of the fuel and thereby reduce the fuel cycle cost, it is necessary to employ materials which have high resistance to void swelling as well as better high temperature mechanical properties. The Indian fast reactor program began with the commissioning of the 40 MWtFast Breeder Test Reactor (FBTR). The core structural material of FBTR is 20% cold worked 316 austenitic stainless steel (SS). For the 5000 MetPrototype Fast Breeder Reactor (PFBR) which is in an advanced stage of construction at Kalpakkam, 20% cold-worked alloy D9 (14Cr-15Ni-Ti SS) has been selected as the cladding and wrapper tube material for the initial core. The target burnup of the fuel is 100 GWd/t. Advanced austenitic stainless steel and oxide dispersion strengthened steels are being developed for achieving fuel burnup higher than 100 GWd/t. An advanced alloy D9 containing controlled amounts of titanium, silicon and phosphorous has been developed. This alloy named as IFAC-1 (Indian Fast Reactor advanced Clad-1) SS is aimed at thermal creep properties comparable to that of alloy D9, and superior void swelling resistance upto a target burn-up of about 150 GWd/t. The nominal chemical composition of IFAC-1 SS is 14Cr-15Ni-.25Ti-.75Si-.04P. The chemical composition has been optimized after extensive evaluation of the tensile, creep and microstructural stability of fifteen laboratory heats with different amounts of titanium, silicon and phosphorous. Void swelling behavior was studied using ion irradiation. IFAC-1 SS contains higher levels of low melting eutectic phase forming elements such as phosphorous, and so is susceptible to solidification cracking. Extensive pulsed TIG welding trials have been carried out on IFAC-1 SS/316LN SS weld joints with varied weld parameters to find out the feasibility of obtaining solidification crack-free welds and the optimum welding parameters have been established. This paper gives an overview of the development of this advanced core structural material for SFRs.

Fuel burnup performance of an OTTO refueling pebble bed reactor with burnable poison loading

Fuel burnup performance has been analyzed for a pebble bed reactor with a once-through-then-out (OTTO) refueling scheme and compared with a reference multi-pass scheme. A new fuel pebble was designed by adding spherical B4C particles into its free fuel zone for controlling the infinite multiplication factor during burnup, and then reducing the axial power peak of the OTTO scheme. The objective is to maximize the fuel burnup performance of the OTTO scheme while keeping the power peak under a limit and ensuring the core criticality. Numerical calculations were performed based on the 400 MWt pebble bed modular reactor (PBMR) using the MVP code. For the fuel pebble of the PBMR containing 9 g uranium with 9.6 wt% 235U enrichment, 1600 B4C particles with a radius of 70 μm are determined to flatten the k∞ curve in the early burnup stage. The dependences of the neutronic properties of the core with the OTTO scheme on target fuel burnup show that the maximum target burnup of 74 GWd/t can be achieved so that the power peak is reduced to about 10.80 W/cm3 which is approximate that of the multi-pass scheme (10.85 W/cm3). This target burnup is about 22% less than that of the multi-pass scheme (95 GWd/t), i.e. the fuel utilization efficiency of the OTTO scheme is about 22% lower, which could be compensated by the construction and operation cost of the fuel handling system. This result also suggests that further investigations of the fuel burnup performance and other properties are needed in both neutronic and thermal hydraulic viewpoints to find out the optimal core performance.

Reactivity considerations for the on-line refuelling of a pebble bed modular reactor—Illustrating safety for the most reactive core fuel load

In the multi-pass fuel management scheme employed for the pebble bed modular reactor the fuel pebbles are re-circulated until they reach the target burn-up. The rate at which fresh fuel is loaded and burned fuel is discharged is a result of the core neutronics cycle analysis but in practice (on the plant) this has to be controlled and managed by the fuel handling and storage system and use of the burnup measurement system. The excess reactivity is the additional reactivity available in the core during operating conditions that is the result of loading a fuel mixture in the core that is more reactive (less burned) than what is required to keep the reactor critical at full power operational conditions. The excess reactivity is balanced by the insertion of the control rods to keep the reactor critical. The excess reactivity allows flexibility in operations, for example to overcome the xenon build up when power is decreased as part of load follow. In order to limit reactivity excursions and to ensure safe shutdown the excess reactivity and thus the insertion depth of the control rods at normal operating conditions has to be managed. One way to do this is by operational procedures. The reactivity effect of long-term operation with the control rods inserted deeper than the design point is investigated and a control rod insertion limit is proposed that will not limit normal operations. The effects of other phenomena that can increase the power defect, such as higher-than-expected fuel temperatures, are also introduced. All of these cases are then evaluated by ensuring cold shutdown is still achievable and where appropriate by reactivity insertion accident analysis. These aspects are investigated on the PBMR 400MW design.

Nuclear safeguards considerations for pebble bed reactors (PBRs)

Recent reports by the Department of Energy National Laboratories have discussed safeguards considerations for low enriched uranium (LEU)-fueled pebble bed reactors (PBRs) and the need for bulk accountancy of the plutonium in “used fuel.” These reports fail to account for the degree of plutonium dilution in the graphitized-carbon pebbles that is sufficient to meet the International Atomic Energy Agency (IAEA) “provisional” guidelines for termination of safeguards on “measured discards.” The thrust of this finding is not to terminate safeguards but to limit the need for specific accountancy of plutonium in stored used fuel. While the residual uranium in the used fuel is not sufficiently diluted to meet the IAEA provisional guidelines for termination of safeguards, the estimated quantities of the uranium minor isotopes 232U and 236U in the used fuel at the target burnup of ∼90 Gigawatt-days per metric ton (GWD/MT) exceed standard specification limits for reprocessed uranium and will require extensive blending with either natural uranium or uranium enrichment tails to dilute the 236U content to fall within specification. Hence, the PBR used fuel is less desirable for commercial reprocessing and reuse than that from light water reactors. Also the PBR specific activity of a reprocessed uranium isotopic mixture and its A2 values for effective dose limits if released in a dispersible form during a transportation accident are more limiting than the equivalent values for light-water-reactor used fuel at 55GWD/MT without accounting for the presence of the principal carry-over fission product (technetium, 99Tc) and plutonium contamination. Thus, the potentially recoverable uranium from PBR used fuel carries reactivity penalties and radiological penalties likely greater than those for reprocessed uranium from light water reactors. These factors impact the economics of reprocessing, but a more significant consideration is that reprocessing technologies for coated particle fuels encased in graphitized carbon have not progressed beyond laboratory-scale demonstrations. However, key equipment that has been tested in the past (such as graphite burners and electrolytic disintegration/dissolution devices) is not listed on either the “Trigger List” or the “Dual Use List” for mandatory export controls. If gross burnup determined from fission-product gamma-ray inspection of a discharged pebble cannot be correlated acceptably with predicted plutonium content of the pebble, development and testing may be required on detector concepts for more directly measuring the plutonium content in a discharged pebble to ensure that its placement in the spent fuel storage tanks is for an acceptable measured discard of diluted plutonium.

Specific Radioactivity of Neutron Induced Radioisotopes: Assessment Methods and Application for Medically Useful 177Lu Production as a Case

The conventional reaction yield evaluation for radioisotope production is not sufficient to set up the optimal conditions for producing radionuclide products of the desired radiochemical quality. Alternatively, the specific radioactivity (SA) assessment, dealing with the relationship between the affecting factors and the inherent properties of the target and impurities, offers a way to optimally perform the irradiation for production of the best quality radioisotopes for various applications, especially for targeting radiopharmaceutical preparation. Neutron-capture characteristics, target impurity, side nuclear reactions, target burn-up and post-irradiation processing/cooling time are the main parameters affecting the SA of the radioisotope product. These parameters have been incorporated into the format of mathematical equations for the reaction yield and SA assessment. As a method demonstration, the SA assessment of 177Lu produced based on two different reactions, 176Lu (n,γ)177Lu and 176Yb (n,γ) 177Yb (β- decay) 177Lu, were performed. The irradiation time required for achieving a maximum yield and maximum SA value was evaluated for production based on the 176Lu (n,γ)177Lu reaction. The effect of several factors (such as elemental Lu and isotopic impurities) on the 177Lu SA degradation was evaluated for production based on the 176Yb (n,γ) 177Yb (β- decay) 177Lu reaction. The method of SA assessment of a mixture of several radioactive sources was developed for the radioisotope produced in a reactor from different targets.