A track detector technique for determining ion source emission profiles

Solid state nuclear track detectors have been used to directly measure for an ion source the beam emission cross sections along the ion beam trajectory. Results are demonstrated for an ion source of the surface ionization type using solid source materials to obtain ¹⁴⁷Sm or ⁶Li isotopes. Cellulose nitrate LR 115 was used as the detector to register the alpha tracks from the radioactive decay of the implanted ¹⁴⁷Sm and from the ⁶Li (n, α) reaction induced by thermal neutrons in a nuclear reactor.     

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.     

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of ²³⁷Np, ²³⁸Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of ²³⁷Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO₂ or UO₂–Gd₂O₃), which was similar to the case of PWR fuel. The differences in the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ¹⁴⁹Sm, and ¹⁵¹Sm have a large impact on k_eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.     

A method for the characterization of fuel failure in operating PWRs

A method is presented for the determination of the number of defective fuel pins in PWRs. The method uses measured values of reactor coolant activity of two fission-product isotopes (e.g. ¹³¹I and ¹³³I) and evaluates the possible number of failures of two different sizes of break. Encouraging results have been obtained for two different reactors which experience a significant amount of failed fuel verified by sipping. The present method can be extended for application to BWRs.     

Irradiation analysis of U, Am samples irradiated in experimental fast reactor “Joyo” for protected plutonium production in fast breeder reactor blanket

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper summarizes analysis of the individual Am and U samples irradiation in Joyo to re-evaluate the results of Pu isotopes in the measure of proliferation resistance, and to combine the results for the prediction of DU-Am irradiation especially in the production of Pu isotopes. By the prediction of DU-Am oxide fuel in fast reactor environment with detail fuel irradiation analysis, it was confirmed that neutron moderation and fuel size affects the produced Pu isotope and its vector due to the very high sensitivity of ²³⁸U resonance capture reaction, the larger diameter fuel is more preferable in the case of moderated neutron spectrum environment for denaturing Pu in fast reactor blanket. Finally proliferation resistance of all the Pu produced in U, Am sample irradiation and DU-Am fuel irradiation prediction were evaluated based on decay heat and spontaneous fission neutron rate, and it was confirmed ²⁴¹Am produces un-attractive Pu to abuse from the beginning to the end of irradiation, and more than 2% of ²⁴¹Am doping is required to enhance the proliferation resistance of Pu to MOX grade and Kessler’s proposal in moderated neutron spectrum environment in fast reactor.     

Comparisons of HELIOS,ORIGEN2, and Monteburns Calculated 241 Am and 243 Am Concentrations to Measured Values for PWR,BWR, and VVER Spent Fuel

A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating ²⁴¹ Am and ²⁴³Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The ²⁴¹Am and ²⁴³Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for ²⁴¹Am and ±20% for ²⁴³Am) and the Garigliano measurements (within ±12% for ²⁴¹ Am and ±20% for ²⁴³Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for ²⁴¹Am and ±12% for ²⁴³Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials.     

ABSTRACTS OF JAPANESE ARTICLES in NIHON-GENSHIRYOKU-GAKKAI SHI

Burn-up calculation and comparison of the results were carried out to clarify the differences among the following latest evaluated nuclear data libraries: JENDL-3.2, ENDF/B-VI and JEF-2.2. The analyses showed that the differences seen among the current evaluated nuclear data libraries are small for evaluation of the amounts of many uranium and plutonium isotopes. However, several nuclides important for evaluation of nuclear fuel cycle as ²³⁸Pu, ²⁴⁴Cm, ¹⁴⁹Sm and ¹³⁴Cs showed large differences among used libraries. The chain analyses for the isotopes were conducted and the reasons for the differences were discussed. Based on the discussion, information of important cross section to obtain better agreement with the experimental results for ²³⁸Pu, ²⁴⁴Cm, ¹⁴⁹Sm and ¹³⁴Cs was shown.     

Effect of Transplutonium Doping on Approach to Long-life Core in Uranium-fueled PWR

The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them ²⁴¹Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides ²⁴²Cm and ²⁴⁴Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile ²³⁹Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of ²⁴¹ Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100GWd/tHM with about 20% ²³⁸Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Analysis of Curium in Mixed Oxide Fuel Irradiated in the Experimental Fast Reactor JOYO for the Evaluation of Its Transmutation Behavior

Curium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Curium was isolated from the irradiated MOX fuel by anion-ex- change chromatography using a mixed medium of nitric acid and methanol. The isotopic ratio of curium and its content were determined by thermal ionization mass spectroscopy and alpha-spectrometry, respectively. The curium content was less than 0.004 at% even at high burnup of 120GWd/t, which is much smaller than that of PWR-MOX at 60 GWd/t. On the basis of present analytical results, the transmutation behavior of curium isotopes in a fast reactor was discussed from various viewpoints. Transmutation rates of curium isotopes were estimated; the rate for ²⁴⁶Cm, which is known to be a key nuclide in the transmutation of curium, was larger than the previously reported value. It was concluded from these evaluations that the fast reactor was suitable for the incineration of curium.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

Performance of the gas turbine-modular helium reactor fuelled with different types of fertile TRISO particles

Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with ²³⁹Pu, ²³³U or ²³⁵U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of ²³²Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes (²³⁹Pu, ²³³U and ²³⁵U), requires a quantity of thorium 25–30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 μm. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, ²³³U, ²³⁵U and ²³⁹Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium.     

Fast Reactor Operating on Uranium–Plutonium Oxide Fuel in a Self-Regulatable Regime

Mathematical simulation is used to show that it is possible to develop a fast reactor operating on uranium–plutonium oxide fuel (UO₂)_1–x (PuO₂) _x , the same for all fuel elements in the core, and with uranium carbide in breeding elements with heavy coolant (PbBi eutectic). A self-regulatable regime is obtained in the reactor. This enhances safety while minimizing control. Tailings uranium with 0.1% ²³⁵U and a mixture of plutonium isotopes, which is obtained from spent fuel, making it possible to conduct operation in an actinide-closed fuel cycle, is used in the fuel and uranium carbide. ²³⁸U is actually consumed in the reactor, but most fission products are produced from ²³⁹Pu.     

Calculation of the time for establishing radiation equivalence of high-level wastes

The potential harm of long-lived radionuclides estimated as the product of the activity of a radionuclide and the dose coefficient is examined. The potential harm of actinides in high-level wastes is calculated taking account of the harm due to their decay products. At the same time, an analogous calculation is performed for the uranium isotopes ²³⁸U, ²³⁵U, and ²³⁴U consumed in the reactor. The value obtained, which depends on the time, can be regarded as the averted harm. The time for establishing radiation equivalence between the high-level wastes and the consumed uranium is determined as the time in which the potential harm from actinides becomes equal to the averted harm. It depends on the holding time of the spent fuel before radiochemical reprocessing. For a 5 yr holding period, it is 49000 yr.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 123–129, February, 2005.     

Application of stable boron isotopes

This article presents a brief survey of the basic applications of stable boron isotopes as materials with an altered isotope composition in cases where the differences between the nuclear properties of boron isotopes are used directly. The neutron flux transformation into heavy ionizing particles by means of the B¹⁰ (n, )Li⁷ reaction and the large value of the effective thermal neutron cross section of this reaction make it possible to use B¹⁰ in medicine, nuclear studies, and radiation chemistry. The differences between the neutron cross sections of B¹⁰ and B¹¹ make It possible to use these isotopes in reactor construction and, in particular, to use B¹⁰ in materials for control rods and reactor shields.     

Effects of Organic Impurities on BWR Water Chemistry, (2)

In order to keep ⁶⁰Co radioactivity low without incurring serious disadvantages regarding integrity of fuel cladding and reliability of structural materials, it is desirable to relate cooling water (especially ⁶⁰Co behavior), structural materials and fuel cladding quantitatively. Organic impurities, such as fragments of cation resin, are fed into the reactor where they decompose, increasing conductivity of the reactor water and decreasing pH, which, in turn, increase ⁶⁰Co radioactivity in the reactor water and then, shutdown dose rate. The relationships between ⁶⁰Co radioactivity in the reactor water and pH have been analyzed by using water chemistry experience at operating BWR plants as well as laboratory data of resin decomposition in the water at elevated temperature, and analytical formulas to explain the shutdown dose rate increase caused by organic impurities intrusion were proposed. Effects of organic impurities on the shutdown dose rate can be moderated preventively by application of weak alkali control.     

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.     

Adapting the deep burn in-core fuel management strategy for the gas turbine – modular helium reactor to a uranium–thorium fuel

In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine – modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium–thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: ²³⁵U, which represents the 20% of the fresh uranium, ²³³U, which is produced by the transmutation of fertile ²³²Th, and ²³⁹Pu, which is produced by the transmutation of fertile ²³⁸U. In order to compensate the depletion of ²³⁵U with the breeding of ²³³U and ²³⁹Pu, the quantity of fertile nuclides must be much larger than that one of ²³⁵U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of ²³⁵U. At the same time, the amount of ²³⁵U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k_eff and mass evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium–thorium fuel.     

Rejuvenation of spent fuel in the force-free helical reactor (FFHR)

This study considers spent fuel rejuvenation potential of the force-free helical reactor (FFHR), which is a relevant heliotron-type D-T fusion reactor. For this purpose, three different spent fuels were selected, i.e.: (1) CANDU, (2) PWR-UO₂, and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the FFHR. The FZ was cooled with high-pressured helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years by 75% plant factor (η) under a first-wall neutron load (P) of 1.5 MW/m². The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δt) of 1/12 year (one month), by using the interface program (code). In all investigated spent fuel cases, the tritium self-sufficiency is maintained for DT fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU reactor after a regeneration time of ∼6 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO₂ and -MOX spent fuels after 40 and 34 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during spent fuel rejuvenation. Consequently, the blanket has high neutronic performance for the rejuvenation of the spent fuels.     

An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the ²⁴¹Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the ²⁴⁰Pu to ²³⁹Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor.