Determination of the Fuel-Assembly Burnup in a Research Reactor by Repeated Short-Time Irradiation Followed by γ Spectrometric Measurements

Experiments performed to determine the absolute fuel burnup in spent fuel assemblies in the IRT research reactor at the Moscow Engineering Physics Institute are described. The method is based on measuring the residual amount of ²³⁵U in the spent fuel asemblies with respect to the activity of the fission product ¹⁴⁰La accumulated in fresh and spent fuel assemblies after they were irradiated for a short time in the reactor core. A fresh fuel assembly with known uranium mass was used as a standard. The neutron flux was monitored using Al + Cu and Al + Co wires placed at the center of the fuel assembly. Small corrections for the difference in the spectrum amd the flux density of the neutrons in fuel assemblies with different uranium content were obtained from the calculations. The burnup of the three fuel assemblies studied was determined to within less than 2%.     

Effect of fission products yield library on the calculation of the nuclide composition of the fuel in sodium-cooled fast reactors

The calculation of the composition of irradiated fuel for different degrees of burnup is a basic problem in the analysis of nuclear-radiological safety of objects holding spent fuel assemblies. The yield of fission products is one of the important initial indicators in burnup calculations. Methods for compiling libraries of fission products yield on the basis of the ENDF/B up-to-date evaluated nuclear data files are described. The nuclide composition of uranium oxide and uranium-plutonium-zirconium metal fuel in sodium-cooled fast reactors is analyzed by means of high-precision calculations performed with different fission product yields libraries using different computer codes MONTEBURNS–MCNP5–ORIGEN2 and the results are presented.     

Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.     

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the ¹³⁴Cs/¹³⁷Cs, Pu, and U isotope ratio measurements.     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

Chemical state of fission products in irradiated uranium carbide fuel

The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.     

Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.     

Method for identifying 17-place casings with spent fuel assemblies of AMB reactors at the Beloyarskaya nuclear power plant

A method is proposed for identifying a cylindrical case with a jacket containing 17 spent fuel assemblies from the AMB reactors at the Beloyarskaya nuclear power plant on the basis of measurements of the radiation characteristics in lateral surface of the jacket opposite the fuel assemblies in the outer and inner rows. Computational validation using the PRIZMA and MCNP computer codes is given for the method. It is shown that the collection of signals from detectors is specific to each jacket, and this makes it possible to use it as an identifying indicator. __________ Translated from Atomnaya énergiya, Vol. 102, No. 6, pp. 363–367, June, 2007.     

Spent fuel burnup estimation by Cerenkov glow intensity measurement

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to the declared burnups for the tested spent fuel assemblies for cooling times ranging from 900-2.000 d     

Method of nondestructive monitoring of fuel-element seal tightness

A method of monitoring the seal tightness of fuel elements is described. The method is based on determining leaks of gaseous fission products during annealing of preirradiated samples. The basic steps using this method are described and mathematical expressions for determining the constants of the activation model of the leakage of gaseous fission products, which is used for processing and analyzing experimental results, are presented. The measurements results obtained using the method develop are compared with the results obtained in reactor tests of spherical fuel elements. An example using the technique is described: simulators, for which the temperature dependence of ¹³⁵Xe leakage (to 1400 K) is obtained and a quantitative relation between the number of microfuel elements with damaged coatings and leakage of gaseous fuel products, are tested. It is found that technological contamination of graphite of the simulators by uranium is present.     

Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

We report the measurement of elastic constants of non-irradiated UO₂, SIMFUEL (simulated spent fuel: UO₂ with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.     

An investigation on recycling the recovered uranium from electro-refining process in a CANDU reactor

Feasibility studies for recycling the recovered uranium from electro-refining process of pyroprocessing into a Canada Deuterium Uranium (CANDU) reactor have been carried out with a source term analysis code ORIGEN-S, a reactor lattice analysis code WIMS-AECL, and a Monte Carlo analysis code MCNPX. The uranium metal can be recovered in a solid cathode during an electro-refining process and has a form of a dendrite phase with about 99.99% expecting recovery purity. Considering some impurities of transuranic (TRU) elements and fission products in the recovered uranium, sensitivity calculations were also performed for the compositions of impurities. For a typical spent PWR fuel of 3.0 wt.% of uranium enrichment, 30 GWD/tU burnup and 10 years cooling, the recovered uranium exhibited an extended burnup up to 14 GWD/tU. And among the several safety parameters, the void reactivity at the equilibrium state was estimated 15 mk. Additionally, a simple sphere model was constructed to analyze surface dose rates with the Monte Carlo calculations. It was found that the recovered uranium from the spent PWR fuel by electro-refining process has a significant radioactivity depending on the impurities such as fission products.     

Study of the behavior of vver and pwr fuel irradiated in the hbwr reactor (halden,norway)

The methods, techniques, and results of comparative studies of VVER and PWR fuel tested in the HBWR reactor (Norway) are presented. Experimental VVER fuel elements with uranium dioxide fuel were fabricated at the Machine Building Plant (MSZ) (in Elektrostal) using standard technology; the experimental PWR fuel elements were fabricated according to the model specifications. The results obtained made possible a comparative evaluation of the changes in the thermophysical parameters and the heat and radiation resistance of the two types of fuel as well as the kinetics of the gaseous fission products as a function of the heat load and burnup.     

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using ¹³⁷Cs and ¹³⁴Cs burnup monitors was employed in different reactors to estimate the amount of ²³⁵U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of ¹³⁴Cs/¹³⁷Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.     

Use of Recovered Uranium and Plutonium in Thermal Reactors

A possible version of the VVER-1000 fuel cycle without separation of uranium and plutonium during reprocessing of spent fuel is examined. In this fuel cycle, the uranium-plutonium regenerate obtained, from which other actinides and fission products have been removed, is used after enriched natural uranium is added for preparing VVER fuel. The results of a calculation of the content of uranium and plutonium isotopes in the spent uranium-plutonium fuel after one and two recycles in VVER-1000 are presented. The main advantages of the fuel cycle are discussed: lower risk of plutonium proliferation, savings of natural uranium, and less spent fuel as compared with an open uranium fuel cycle. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 136–141, August 2005.     

Nuclide separation from spent nuclear fuels by using tertiary pyridine resin

The newly nuclide separation system from spent nuclear fuels is proposed. The proposed separation system consists of recovery of nuclear fuel elements, separation of trivalent minor actinide from lanthanide, and separation of some fission products such as strontium. This separation system is based on the chromatographic technique using the tertiary pyridine resin. Separation experiments using mixed oxide fuel highly irradiated in fast reactor “Joyo” were carried out. The recovery of plutonium, the separation of minor actinide from fission products including lanthanides, and the separation of americium and curium were achieved. The recovery or removal of platinum group elements and technetium was also investigated, and the removal of these elements prior to the main reprocessing process has been proposed.     

Growth of the interaction layer around fuel particles in dispersion fuel

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl_x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.     

Environmentally acceptable nuclear fuel cycle development of a new reprocessing system

The aim of the present study is to establish a new reprocessing system for spent nuclear fuel, which would overcome the environmental problems in the nuclear fuel cycle. In order to achieve this, the following subjects have to be conquered: recoveries of high ratios of uranium and trans uranium elements from spent nuclear fuel, separations of strong radioactive elements, such as Sr and Cs, and assurance of the extreme safety during operation. The last subjects might be of particular importance in order to avoid any potential danger. Therefore, in the present system all processes were performed under mild aqueous conditions. Experiments were carried out for a simulated spent fuel solution, which was calculated from the ORIGEN CODE containing uranium and 17 major elements. The system consists of the following processes: 1. dissolution of spent UO₂ fuel involving off-gas treatment of I and Ru; 2. neutralization of the dissolved fuel solution with NaHCO₃---Na₂CO₃ mixed solution to be slightly basic at pH about 9 followed by the separation of precipitated fission products by centrifugation; 3. separation of Cs by a precipitation method using tetraphenylborate ion; 4. recovery of U, Np and Pu as precipitates of hydrolyzed compounds from alkaline solution; 5. separation of Am and Cm from lanthanide elements. The concentration of residual uranium in the final solution was measured to be ppm order, indicating that the decontamination factor of U was as large as 10⁴. Other hexa-valent actinide ions, NpO₂²⁺ and PuO₂²⁺, also have extremely large stability constants for the complex formation with carbonate ion, and are expected to behave similarly with UO₂²⁺. In conclusion, the present reprocessing system enables us to recover U, Pu and Np from spent nuclear fuel by means of a simple precipitation method in much higher ratios compared with other reprocessing methods, to separate hazardous Cs and Sr from high-level waste, and to exclude any potential danger owing to chemical processes under mild aqueous conditions.     

Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.     

Implementation of burn-up credit approach for transport and storage cask

Abstract

TN International currently uses burn-up credit methodology for the design of casks dedicated to the transport of pressurised water reactor uranium oxide spent fuel assemblies. As long as the fuel enrichment of the pressurised water reactor fuel assemblies was sufficiently low, a burn-up credit methodology based on the sole consideration of actinides and the use of a partial burn-up was satisfactory to cover the needs without necessity to design new casks. Nevertheless, the continuous increase in the fuel enrichment during the last decade has led TN International to continue the investigations on the burn-up credit methodology to limit both the increase in the neutron poison content in the new basket designs and the burn-up constraints attached to the acceptability of the fuel assemblies for transport. The strategy of TN International was then to take benefit of the large negative reactivity reserves, which might be gained by the consideration of the fission products coming from the fuel irradiation. A big step forward has recently been reached by TN International on this field with the definition of an advanced burn-up credit methodology based on the consideration of relevant fission products recommended by OECD. In the meantime, TN International has taken the opportunity to use such burn-up credit approach in the design of the TN 24 E transport and storage cask developed for the German nuclear power plants. The relevant task has been carried out according to the German standard DIN 25712 for burn-up credit application. The present paper will describe the basic principles of the burn-up credit methodology implemented by TN International such as:

(i) the current state of the art concerning the burn-up credit application in the criticality assessment

(ii) the basic approach used for the implementation of the advanced burn-up credit methodology (bounding axial burn-up profiles, fuel irradiation parameters, fission products, etc.)

(iii) the area of validity of the TN International burn-up credit approach with fission products

(iv) example of application of the burn-up credit methodology for the design of the TN 24 E transport and storage cask under licensing in Germany

(v) the perspectives of development of the burn-up credit methodology.