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.     

Plutonium build-up credits for a material test research reactor and influence of cross-section differences on actinide production

Burnup calculations with SARC system were carried out to analyse the effects of plutonium build-up on criticality of MTR type research reactor PARR-1 using several WIMSD libraries based on evaluated nuclear data files ENDFB-VI.8, JEF-2.2, JEFF-3.1 and JENDL-3.2. For equilibrium core of the reactor, it was found that a net reactivity of more than 3.5 mk is induced due to build-up of plutonium isotopes during depletion. The plutonium credit amounts to 3% of the length of equilibrium cycle. From the analysis of actinide production in the core during burnup, it was observed that in most of the cases, the amounts of actinides obtained using various cross section libraries agree fairly with each other, however, significant differences were observed for ²³⁸Pu, ²⁴¹Pu, ^242mAm, ²⁴³Am, ²⁴²Cm and ²⁴⁴Cm for some libraries. The actinide chain analysis was conducted to investigate the reasons for the observed differences.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.     

Correlation of Amounts of Transplutonium Nuclides with Burnup in the JPDR-I Spent Fuel

The amounts of ²⁴¹Am, ²⁴²Cm and ²⁴⁴Cm were determined by means of a radiochemical technique in several specimens taken from the spent fuel of the JPDR-I. The yields of the transplutonium nuclides were examined in connection with the burnup determined by a conventional method. It was found that the burnup correlates well with the yield ratios of various transplutonium nuclides.     

Fuel irradiation of the first batches produced for the Chinese HTR-10

An irradiation test of four spherical fuel elements (SFE) had been performed in the Russian reactor IVV-2M. The elements were sampled randomly from the first and second product batches which were manufactured for the 10 MW high-temperature gas-cooled test reactor (HTR-10). The maximum burnup of the irradiated fuel elements reached 107,000 MWd/tU and the maximum fast neutron fluence was 1.31 × 10²⁵ m⁻². The release-to-birth rate ratio (R/B) did not increase significantly during irradiation. However, an in-pile heating-up test of element SFE 7 in Capsule 5 led to a failure of approximately 6% of the coated particles. After the test it was estimated that the fuel temperature had very likely been much higher than the intended 1600 °C.     

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.     

Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.     

Benchmark of Advanced Burner Test Reactor model using MCNPX 2.6.0 and ERANOS 2.1

Significant research is currently being performed whereby fast reactor cores have been designed to burn transuranic materials reducing the volume and long-term radiotoxicity of spent nuclear fuel. These core and depletion models depend on various computer codes. This research used MCNPX 2.6.0 and ERANOS 2.1 to model a standard 250 MW Advanced Burner Test Reactor (ABTR) core. The intent was to benchmark criticality and burnup results from a stochastic Monte Carlo code and a deterministic depletion code using a standard ABTR model created by Argonne National Laboratory. Because each of these codes solves the transport and burnup problem differently, there is a need to benchmark the core models in order to verify results and identify root causes for significant differences in results between codes. Flux calculations in ERANOS were performed using diffusion theory, Legendre polynomial approximations (using the VARIANT module) and discrete ordinates methods. The k-effective for the higher order transport models remained within 1000 pcm of the MCNPX model. The difference between the total heavy nuclide mass balance in ERANOS using the various flux calculations and the MCNPX depletion model was less than 0.4% out to a burnup of 1095 days (67.45 GWd/MTHM). The percent delta between the codes as a fraction of the fissioned mass was 1.34%. For the isotopes with large concentrations, such as ²³⁸U and ²³⁹Pu, the mass differences were 0.38% and 0.01% respectively. The mass difference for ²⁴¹Am was also small at 0.42%. Notable isotopes in small quantities with larger mass differences were ²⁴²Am, ²⁴²Cm, ²⁴³Cm and ²⁴⁶Cm where differences ranged from 0.1 to 0.2% after 26 days and increased to 11–136% at 1095 days.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

PWR（U）乏燃料中超铀元素在混合堆快裂变包层内嬗变研究

从中子学角度对ＰＷＲ（Ｕ）乏燃料中的超铀元素（２３８Ｐｕ，２３９Ｐｕ，２４１Ｐｕ，２４１Ａｍ，２４３Ａｍ，２３７Ｎｐ，２４４Ｃｍ）在聚变－裂变混合堆快裂变包层内嬗变的可行性进了研究。利用一维中子输运和燃耗计算程序ＢＩＤＥＣＡＹ译不同燃料组分的四个快裂变包层进行分析计算。结果表明，在聚变－裂变混合堆快裂变包层内安全，高效地嬗变ＰＷＲ（Ｕ）乏燃料中的超铀元素是可能的。     

JEFF-3.1.1 Nuclear Data Validation for Sodium Fast Reactors

The JEFF-3.1.1 Nuclear Data Library is the latest version of the Joint Evaluated Fission and Fusion Library. We present the status of the validation of this library using the Monte Carlo Code TRIPOLI 4.5 and the deterministic code package ERANOS 2.2 for fast reactor calculations. For this purpose, we reanalyze a selected set of integral experiments performed in the MASURCA mock-up (CEA/CADARACHE), in the ZPPR mock-up at ANL (USA), and in the SUPERPHENIX Power Reactor. Furthermore, we also present the analysis of pure sample irradiation experiments PROFIL and PROFIL-2 performed in the PHENIX reactor, as this kind of experiment provides a direct feedback on nuclear capture data. We observe good performances of these calculation tools for criticality calculations and fuel inventory prediction. From this validation work, some required improvements on nuclear data are highlighted, as well as the need for new specific integral experiments. The main trends observed are the following:

—Reactivity of clean and fresh cores: the results obtained with JEFF-3.1.1 are consistent with those obtained using JEFF-3.1 within 80 pcm, but there is an overestimation of the calculated reactivity of all the experiments between 40 and 800 pcm depending on the spectrum hardness (Pu content) and fuel composition, the discrepancy being larger in hard spectra. Additional investigations are in progress to understand this behaviour.

—Integral capture cross sections (PROFIL and PROFIL-2):/C/E - 1/≤ 3%, except for ²⁴¹Pu (C/E ≈ 1:08), ²⁴²Pu (C/E ≈ 1:18), ²³⁷Np (C/E ≈ 0:92), ²⁴³Am (C/E ≈ 0:93), ²⁴⁴Cm (C/E ≈ 1:35); impact of the trends observed on the individual fission products put in PROFIL and PROFIL-2 is ≈ 4% on the part of Δ_ρburnup due to fission products.     

Specific activity of243Am and243Cm in the fuel of the 4th power-generating unit of the Chernobyl nuclear power plant

The activity ratios²⁴¹Am/²⁴¹Am.²⁴³Cm/²⁴⁴Cm, and²⁴²Cm/²⁴⁴Cm in core samples taken at the industrial site of the object “Cover” were measured. The content of²⁴³Am and²⁴³Cm in the fuel in the 4th power-generating unit of the Chernobyl nuclear power plant was estimated on the basis of an analysis of these data. The experimental data were compared with the theoretical calculations. 2 figures, 9 references. Scientific Center “Institute of Nuclear Studies,” Ukrainian National Academy of Sciences. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 327–329, November, 1999.     

Melting Temperature of Irradiated UO2 and UO2-2wt%Gd2O3 Fuel Pellets up to Burnup of about 30 GWd/tU

Melting temperature of UO₂ and UO₂-2w/oGd₂O₃ fuels irradiated in a commercial LWR were determined by a thermal arrest technique in a burnup range up to approximately 30GWd/tU.

No decrease in the melting temperature was observed on both UO₂ and UO₂-2w/oGd₂O₃ fuels with increment of burnup to 30GWd/tU. It was also found that the Gd₂O₃ addition below 2w/o has no influence on the melting temperature.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

In the 1980s, a series of integral experiments was conducted in FCA-IX assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, ²³⁷Np, ²³⁸Pu, ²³⁹Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm. Regarding the fission rate ratios relative to ²³⁹Pu, benchmark models had been recently developed for validation of nuclear data for the TRU's fission cross sections. In this paper, the latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, are compared on the benchmark models. For the libraries, the analyses by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of ²⁴⁴Cm to ²³⁹Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of ²³⁸Pu to ²³⁹Pu measured in the intermediate neutron spectrum. The causes of discrepancies are furthermore clarified by sensitivity analyses.     

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.     

Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code

Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4–7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0-31 MWd/kgUO₂. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release.

In the irradiation period up to about 23 MWd/kgUO₂, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO₂ pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO₂ cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.     

Fuel management of the HTR-10 including the equilibrium state and the running-in phase

The mode of fuel management of the HTR-10 was studied, including the simulation of the fuel shuffling process and the measurement of the burnup of a fuel element. The prior consideration was the design of the equilibrium state. Based on this the fuel loading of the initial core and the fuel shuffling mode from the initial core through the running-in phase into the equilibrium state were studied. The code system VSOP was used for the physical layout of the HTR-10 at the equilibrium state and in the running-in phase. For the equilibrium state, in order to lessen the difference between the peak and the average burnup, 5-fuel-passage-through-the-core was chosen for the fuel management. The average burnup of the spent fuel for the equilibrium core is 80 000 MWd t⁻¹, and the peak value of it is less than 100 000 MWd t⁻¹ when the burnup of the recycled fuel element is under 72 000 MWd t⁻¹. The mixture of fuel element and graphite element was used for the initial core loading, the volume fractions of the fuel and the graphite elements were 0.57 and 0.43, respectively. During the running-in phase, the volume fraction of graphite will decrease with the fresh fuel elements being loaded from the top of the core and the graphite elements discharged from the bottom of the core. The fuel shuffling mode is similar to that of the equilibrium state. The burnup limit of recycled fuel element is also 72 000 MWd t⁻¹ and the peak burnup is less than 100 000 MWd t⁻¹. Finally the core will be full of fuel elements with a certain profile of burnup and reaches the equilibrium state. According to the characteristics of the pebble-bed high temperature gas-cooled reactor, a calibrating method of concentration of ¹³⁷Cs was proposed for the measurement of fuel burnup.     

244Cm Transmutation in Accelerator-Driven System

Mixed oxide (MOX) fueled LWRs are characterized by enhanced accumulation of curium isotopes compared to uranium fueled LWRs. Toxicity of relatively short-lived isotope ²⁴⁴Cm (T ₁/2=18 yr) appears to be dominant in transuranics discharge. The present paper deals with analysis of the scientific feasibility to transmute ²⁴⁴Cm. Liquid bismuth-curium core operated in a subcritical mode with neutron support from accelerator is proposed to provide transmutation rate through fissioning that exceeds the rate of natural decay by one order of magnitude. Beam current in the range 30–40 mA is required to drive the core. One accelerator-driven transmutor could solve the problem of curium accumulation in a large-scale nuclear energy system based on MOX fueled LWRs.     

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.