Dependence of Fast Reactor Fuel Burnup Characteristics on Nuclear Data Libraries

In the framework of the development of burnup calculation method for commercial fast reactors, a sensitivity analysis was carried out to clarify the dependence of fuel burnup characteristics on nuclear data libraries (NDLs). The following NDLs were compared: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2 and JENDL-3.3. The NDL-dependence of material balance for main heavy metal nuclides (²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu) was small, since the number densities at the end of one-cycle burnup did not change over 1 or 2% among the above-mentioned libraries. Relatively large differences were found for minor actinide nuclides, especially for ²³⁶U, ²³⁷Np, ^242mAm, ²⁴³Am and curium isotopes. The number densities for these nuclides after burning up showed remarkable NDL-dependence over 5% through 50%. A burnup sensitivity analysis system based on the generalized perturbation theory enabled us to find out quantitatively the causative nuclides and reactions, as well as their energy regions.     

Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel

As part of a validation study of burnup calculations of BWR cores, lattice physics analyses were performed on burnups and isotopic compositions of U, Pu and fission product nuclides measured on five samples taken from 9 × 9 BWR fuel assemblies. Burnup calculations in infinite assembly geometry were carried out using MVP-BURN and SRAC codes coupled with major nuclear data libraries. The burnups determined based on the Nd-148 method were from 27.9 to 64.2 GWd/t. The typical relative differences in isotopic compositions (atom/Total-U) between the burnup calculations and measurements were ?2 ～ 19% for ²³⁴U, ?20 ～ 3% for ²³⁵U, ?1.5 ～ 0.1% for ²³⁶U, ?0.04 ～ 0.02% for ²³⁸U, ?4 ～ 11% for ²³⁸Pu, ?11 ～ ?2% for ²³⁹Pu, ?3 ～ 0% for ²⁴⁰Pu, ?12 ～ ?2% for ²⁴¹Pu and ?2 ～ 3% for ²⁴²Pu. They were ?2 ～ 2% for Nd isotopes, ?15 ～ 7% for Eu isotopes, ?13 ～ 1% for Cs isotopes, ?13 ～ 8% for Sm isotopes, 0 ～ 7% for ¹⁴⁷Pm, ?7 ～ ?2% for ⁹⁵Mo, ?2 ～ ?1% for¹⁰¹Ru and 0 ～ 4% for ¹⁰³Rh.     

我国环境土壤中的重要超铀核素分布

Pu、Am、Np是3种重要的超铀核素,环境中的这些核素主要来源于人类的核活动,包括大气层核武器试验、核设施排放和核事故释放等。这些超铀核素不仅具有放射性,还兼具化学毒性。我国地域辽阔,环境土壤类型丰富,在当前核电事业蓬勃发展的背景下,建立和扩大我国环境土壤中这些重要超铀核素的“准本底”数据库是辐射环境安全评价的重要组成部分,也是公众关心的热点问题。近30多年来,研究人员对我国不同环境土壤中这几种超铀核素从不同科学角度开展了调查测量研究。本文对此进行整理和分析,对我国环境土壤中这些重要超铀核素（主要是Pu核素,还包括241Am和237Np）的来源、浓度水平和分布特征进行讨论和综述,为辐射环境安全评价奠定基础。     

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.     

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.     

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.     

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.     

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO₂ assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of ²³⁵U, well reproduce those of ²³⁹Pu and ²⁴⁰Pu, yet underestimate those of ²³⁶U, ²³⁷Nd, ²³⁸Pu, ²⁴¹Pu, and ²⁴²Pu. The inventories of minor actinides are underestimated by the present analysis except for ²⁴¹Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO₂ assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO₂ assemblies were well reproduced by the burnup calculations of MVP-BURN.     

Summation calculations of fission-product decay heat, their uncertainties and their applications to a fast breeder reactor

Prediction of unmeasured fission parameters is described concerning mass yields of fission products, average total kinetic energies of fission fragments and average numbers of prompt neutrons in neutron-induced fission of several nuclides related to nuclear reactors. Reliability of predicting unmeasured mass yields of the products from measured mass yields of the primary fragments is discussed. The predicted values of the mass yields of the products are presented for the fissions of ^233,235U, ^239,241Pu by thermal neutrons, of ²³²Th by 1.38-MeV neutrons and of ²³¹Pa, ²³⁸U and ²³⁷Np by fission-spectrum neutrons.A semi-empirical method based on the statistical theory of nuclear fission is extended to unmeasured distributions of the fragments. The mass yields of the products, the average total kinetic energies of the fragments and the average numbers of prompt neutrons are predicted for the fissions of ²³²U and ²³⁸Pu by thermal neutrons, and of ^234,236U, ²⁴⁰Pu and ²⁴²Pu by 2-MeV neutrons. Errors in the predicted values are also discussed.     

Enhancing BWR proliferation resistance fuel with minor actinides

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the ²³⁸Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides (²³⁷Np and ²⁴¹Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of ²³⁸Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO₂ fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm³) to the top (0.35 g/cm³) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides (²³⁷Np and/or ²⁴¹Am), significantly increases the ²³⁸Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.     

Fission Product Model for BWR Lattice Calculation Code

A benchmark calculation of full fission product was performed for thermal reactor application using an isotope transmutation code DCHAIN based on 185 nuclides with revised nuclear data library. The fission product model for BWR lattice calculation was studied and tested with the benchmark results, and a model containing 45 explicit nuclides and one pseudo nuclide was selected as a reasonably best model to predict the burn up reactivity with high precision for practically all types of fuel and reactor operating conditions. The evaluated thermal cross section and resonance integral for the pseudo nuclide are σ_2,200 = 2.6b and.RI = 10.6b, combined with the pseudo fission yield values of 1.3898, 1.3233, 1.3675 and 1.2773 for fissions from ²³⁵U, ²³⁸U, ²³⁹Pu and ²⁴¹Pu, respectively. The present results are believed as equally applicable to PWR lattice calculation.     

Analysis and application of heavy isotopes in the environment

A growing number of AMS laboratories are pursuing applications of actinides. We discuss the basic requirements of the AMS technique of heavy (i.e., above ～150 amu) isotopes, present the setup at the Vienna Environmental Research Accelerator (VERA) which is especially well suited for the isotope ²³⁶U, and give a comparison with other AMS facilities. Special emphasis will be put on elaborating the effective detection limits for environmental samples with respect to other mass spectrometric methods.At VERA, we have carried out measurements for radiation protection and environmental monitoring (²³⁶U, ^{239,240,241,242,244}Pu), astrophysics (¹⁸²Hf, ²³⁶U, ²⁴⁴Pu, ²⁴⁷Cm), nuclear physics, and a search for long-lived super-heavy elements (Z > 100). We are pursuing the environmental distribution of ²³⁶U, as a basis for geological applications of natural ²³⁶U.     

Sensitivity of charged particle activation analysis for long-lived radioactive nuclide determination

ABSTRACT

Charged Particle Activation Analysis (CPAA) utilizing an 8-MeV proton beam has been studied for determination of 35 long-lived radioactive nuclides. We accumulated the reaction cross section and nuclear decay data by referring to nuclear database supplied by National Nuclear Data Center in Brookhaven National Laboratory. We also calculated the reaction cross sections by using statistical model code ALICE. By using the nuclear data, we have derived determination sensitivity of the radioactive nuclides relative to unit weight and specific radioactivity. The result indicates that several hardly measurable nuclides with long half-lives such as ¹³⁵Cs, ²⁴⁴Pu, ¹²⁹I, ¹²⁶Sn, ⁹³Mo, ¹⁰⁷Pd, ²³⁶U, ²⁴⁸Cm, and ²³⁷Np have high sensitivity. It may be concluded that CPAA can be applied to determination of several long-lived nuclei and will provide a quick and non-destructive analysis method.     

Burning transuranium isotopes in thermal and fast reactors

Energy production in nuclear power plants on the basis of fission processes lead inevitably to fission products and to the generation of new actinide isotopes. Most of these fission products are rather shortlived and decay within less than about 500 years to stable nuclides. However, a few of them, e.g. ⁹⁹Tc and ¹²⁹I, are longlived and may contribute to the radiotoxicity and hazard associated with an envisaged repository for their long-term disposal in a stable geologic formation, e.g. a salt dome. The majority of the generated actinide isotopes are fairly longlived, e.g. ²³⁹Pu with a halflife of more than 20 000 years. Therefore, their direct storage poses a heavy burden on the capacity and the possible environment impact of a repository. Furthermore, the energy content of these actinides could be deployed for producing additional nuclear fission energy after recovering them from unloaded irradiated fuel by suitable reprocessing techniques. Various possibilities exist for burning these actinides in different types of reactors, e.g. in light water reactors (LWRs), or LMFRs, adhering to available technology, or in actinide burners particularly designed for the purpose of their efficient incineration. The different options will be discussed in the paper. Transmutation of the manmade actinides and longlived fission products will require advanced technologies e.g. regarding reprocessing losses, remote fabrication techniques, and most probably, isotope separation processes. However, the almost complete elimination of these nuclides resulting from fission energy production in a continued recycling process may be the only feasible way to limit the effects of nuclear power generation to a tolerable and fair level for generations to come.     

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.     

Plutonium credit in the low enriched uranium fuel of PARR-1

The amount of plutonium (Pu) isotopes and the resultant savings of ²³⁵U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations.The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 × 10⁻³Δk/k. Amount of ²³⁵U equivalent to this value of reactivity was found to be 15.58 ± 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box.     

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.     

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.     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

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.