Determination of within-rod caesium and europium isotopic distributions in high burnup fuel rods through computerised gamma-ray emission tomography

Computerised gamma-ray emission tomography has been applied to single PWR UO₂ fuel rods, with pellet averaged burnups of 52, 71, 91 and 126 GWd/t respectively, for the determination of ¹³⁴Cs, ¹³⁷Cs and ¹⁵⁴Eu internal radial distributions. State-of-the-art image reconstruction techniques, analytical and iterative, have been applied, evaluated and compared using test phantoms first and, in a second step, the actual measured data. Further, linear attenuation maps, previously derived on the same samples by means of gamma-ray transmission tomography, have been used to correct for density inhomogeneities. The final results have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (periphery-to-centre ratios of ∼2.5 for ¹³⁷Cs and ∼3 for ¹³⁴Cs). In addition, a difference in the relative activity distributions of ¹³⁷Cs and ¹³⁴Cs has been observed for all the samples. In contrast, the europium shows an almost flat distribution.     

Usefulness of Single-Crystal Bismuth and Silicon for Neutron Radiography Facility

The atom ratios of Pu/U, ¹³⁴CS/¹³⁷CS and ¹⁵⁴Eu/¹³⁷CS of all the spent fuel assemblies in the full-core of JPDR-1 were calculated. These results were examined through comparison with the values measured by nondestructive γ-ray spectrometry. There were some differences between the calculated and the measured atom ratios of Pu/U and burnup. The calculated atom ratios of ¹³⁴CS/¹³⁷CS and ¹⁵⁴Eu/¹³⁷Cs were slightly less than the measured values for almost all the fuel assemblies. The most probable production amount of Pu, estimated on the basis of the calculated and the measured atom ratios and their accuracies, agreed well with the amount recovered from the reprocessed fuel assemblies.     

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.     

10MW高温气冷堆的燃耗测量研究

10MW高温气冷堆的燃耗测量系统是采用非破坏性高纯锗γ谱仪在线监测来确定燃耗值,利用MCNP4A程序对测量系统的衰减因子进行计算,基于核燃料裂变核素的γ射线能谱分析,以137Cs和134Cs核素活度作为测量对象,并对燃耗测量结果进行讨论.     

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.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

通过分析放射性核素活度鉴别燃料初始富集度的可行性研究

高温气冷堆核电站示范工程（HTR-PM）的反应堆在达到平衡状态前要经过一个较长时间的过渡过程。该过程中堆芯将装入两类燃料球,它们在设计上只有燃料初始富集度不同。反应堆运行要求在过渡过程中要鉴别出装有低富集度燃料的燃料球,并按其燃耗水平不同将其卸出。本文针对此问题,讨论了通过分析燃料球中放射性核素活度（或其比值）以鉴别两类燃料球的方法。堆物理分析软件和KORIGEN软件针对过渡过程的计算结果初步肯定了该方法的理论可行性,并可看出最有可能的鉴别指征量是¹³⁴Cs活度、¹²⁵Sb与¹³⁷Cs的活度比值和¹³⁴Cs活度与¹³⁷Cs活度平方的比值。     

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

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

压水堆核电站乏燃料元件γ扫描测量

在压水堆核电站乏燃料元件检验中，完成了4根完整元件棒、4根破损元件棒的γ扫描测量，元件燃耗分布在9600～45000 MW•d/t（U）之间，获得了完整元件轴向相对燃耗分布、破损元件¹³⁷Cs分布及迁移流失情况。结果显示，破损元件均存在不同程度的Cs迁移流失，破口处存在¹³⁷Cs计数突变（降低）。破损元件¹³⁴Cs/¹³⁷Cs原子比分布与相邻完整元件基本一致，表明¹³⁴Cs、¹³⁷Cs流失比例近似相等，可用¹³⁴Cs/¹³⁷Cs原子比表征其相对燃耗分布；破口处可通过低挥发性核素¹⁵⁴Eu计数水平判断燃料芯块是否缺失。检验结果可为燃料元件破损原因分析及堆内行为分析提供重要依据。     

Comparison between HELIOS calculations and a PWR cell benchmark for actinides transmutation

This paper shows a comparison between the results obtained with the HELIOS code and other similar codes used in the international community, with respect to the transmutation of actinides. To do this, the international benchmark: “Calculations of Different Transmutation Concepts” of the Nuclear Energy Agency is analyzed. In this benchmark, two types of cells are analyzed: a small cell corresponding to a standard pressurized water reactor (PWR), and a wide cell corresponding to a highly moderated PWR. Two types of discharge burnup are considered: 33 GWd/tHM and 50 GWd/tHM. The following results are analyzed: the neutron multiplication factor as a function of burnup, the atomic density of the principal actinide isotopes, the radioactivity of selected actinides at reactor shutdown and cooling times from 7 until 50,000 years, the void reactivity and the Doppler reactivity. The results are compared with the following codes: KAPROS/KARBUS (FZK, Germany), SRAC95 (JAERI, Japan), TRIFON (ITTEP, Russian Federation) and WIMS (IPPE, Russian Federation). For the neutron multiplication factor, the results obtained with HELIOS show a difference of around 1% δk/k. For the isotopic concentrations: ²⁴¹Pu, ²⁴²Pu, and ^242mAm, the results of all the institutions present a difference that increases at higher burnup; for the case of ²³⁷Np, the results of FZK diverges from the other results as the burnup increases. Regarding the activity, the difference of the results is acceptable, except for the case of ²⁴¹Pu. For the Doppler coefficient, the results are acceptable, except for the cells with high moderation. In the case of the void coefficient, the difference of the results increases at higher void fractions, being the highest at 95%. In summary, for the PWR benchmark, the results obtained with HELIOS agree reasonably well within the limits of the multiple plutonium recycling established by the NEA working party on plutonium fuels and innovative fuel cycles (WPPR).     

Radioactivity of fission product and heavy nuclides deposited on soil in Fukushima Dai-Ichi Nuclear Power Plant accident

Based on periodically performed radioactivity measurements on soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, activity ratios to ¹³⁷Cs of fission product and heavy nuclides were obtained for Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Ba, La, Pu, Am, and Cm isotopes. By exponentially fitting or averaging, the activity ratios at the core shutdown were estimated. Using correlations of activity ratios of ¹³⁴Cs to ¹³⁷Cs, and ²³⁸Pu to the sum of ²³⁹Pu and ²⁴⁰Pu against fuel burnup, burnup of the fuel sourcing the deposited activity of the soil was estimated. The activity ratios to ¹³⁷Cs of each nuclide on the deposited activity were divided by those calculated on the fuel at the shutdown to obtain the deposited activity fraction of each nuclide as a relative value to ¹³⁷Cs, which also corresponds to the deposited fraction of each element as a relative value to Cs. The obtained deposited fractions relative to Cs are the orders of 10^?4 to 10^?2 for Sr, 10^?5 to 10^?3 for Nb, 10^?2 to 10^?1 for Mo, 1 to 10 for I, 10^?3 to 10^?2 for Ba, 10^?2 for La, 10^?6 to 10^?3 for Pu, 10^?6 to 10^?4 for Am, and 10^?7 to 10^?5 for Cm. The deposited fractions for Tc, Ag, and Te were not estimated due to the lack of the calculated inventories in the fuel for the relevant measured radioactive nuclides.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

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.     

Isotope correlation technique for Pu mass analysis of PWR UO2 spent fuel rods

A simple and fast method of nuclear material accountancy of pressurized water reactor (PWR) UO₂ spent fuel rods for safeguards application was developed utilizing the isotope correlation between the amounts of ¹³⁷Cs and total Pu. To this end, the following steps were taken: (1) as much destructive analysis (DA) data as possible for segments taken from a PWR UO₂ spent fuel rod were aggregated from publicly available data sources; (2) the DA data were corrected so as to have the same cooling time (i.e., CT = 0 y) and analyzed for outliers; (3) an equation converting the ¹³⁷Cs amount to the Pu amount was obtained by regression analysis with logarithmic curve fitting; and (4) the error in determining the Pu amount was evaluated for the imposition of a limit on the range of burnup (BU) or initial enrichment (IE). It was found that the averaged % error in calibration was determined to be 3.88% ± 2.68% (= mean ± 1 standard deviation) for the BU range over 30 GWd/tU and falling with increasing BU range. On the other hand, there was no benefit in applying the limit of the IE range. Lastly, the Pu-mass difference between various methods was compared and it was found that the difference can be incurred up to 11.4%, according to the choice of method. In conclusion, the proposed isotope correlation technique could be used for input material accountancy with reasonable uncertainty.     

Nondestructive burnup measurements by gamma-ray spectroscopy on spent fuel elements of the RP-10 research reactor

Gamma-ray spectroscopy is an important nondestructive method for the qualification of irradiated nuclear fuels. Regarding research reactors, the main parameter required in the scope of such qualification is the average burnup of spent fuel elements. This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using ¹³⁷Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%.     

Water hyacinth for phytoremediation of radioactive waste simulate contaminated with cesium and cobalt radionuclides

Phytoremediation is based on the capability of plants to remove hazardous contaminants present in the environment. This study aimed to demonstrate some factors controlling the phytoremediation efficiency of live floating plant, water hyacinth (Eichhornia crassipes), towards the effluents contaminated with ¹³⁷Cs and/or ⁶⁰Co. Cesium has unknown vital biological role for plant while cobalt is one of the essential trace elements required for plant. The main idea of this work i.e. using undesirable species, water hyacinth, in purification of radiocontaminated aqueous solutions has been receiving much attention. The controlling factors such as radioactivity concentration, pH values, the amount of biomass and the light were studied. The uptake rate of radiocesium from the simulated waste solution is inversely proportional to the initial activity content and directly proportional to the increase in mass of plant and sunlight exposure. A spiked solution of pH ≈ 4.9 was found to be the suitable medium for the treatment process. The uptake efficiency of ¹³⁷Cs present with ⁶⁰Co in mixed solution was higher than if it was present separately. On the contrary, uptake of ⁶⁰Co is affected negatively by the presence of ¹³⁷Cs in their mixed solution. Sunlight is the most required factor for the plant vitality and radiation resistance. The results of the present study indicated that water hyacinth may be a potential candidate plant of high concentration ratios (CR) for phytoremediation of radionuclides such as ¹³⁷Cs and ⁶⁰Co.     

Distribution of the 134Cs/137Cs ratio around the Fukushima Daiichi nuclear power plant using an unmanned helicopter radiation monitoring system

Large amounts of radioactive substances were released into the environment by the Fukushima Daiichi nuclear power plant (FDNPP) accident. Several research institutes have mapped the distribution of nuclides with long half-lives, such as ¹³⁴Cs and ¹³⁷Cs. Although the ratio of ¹³⁴Cs and ¹³⁷Cs has been believed to be equal without depending on the location of the contaminated area, several researchers report that it is different depending on places quite a little. We measured the energy spectrum of gamma rays in high resolution within an approximately 3-km radius of the FDNPP by using an unmanned helicopter equipped with a LaBr₃(Ce) scintillation detector. Then, we analyzed the ¹³⁴Cs/¹³⁷Cs ratio in the area from these measured data in detail. The results show that the ¹³⁴Cs/¹³⁷Cs ratio is different between the plume trace extending north and the other plume traces. We have obtained valuable data for identification of which radioactive substances were released by individual reactor units.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.