Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO₂ spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO₂ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO₂.     

Photochemically-Induced Valency Adjustment of Plutonium and Neptunium in Nitric Acid Solution Using Mercury Lamp

A photochemically-induced valency adjustment method has been studied to remove Np from the mixed nitric acid solutions of Pu and Np in connection with the Purex reprocessing. The valencies of Pu and Np ions were adjusted to be Pu(HI) and Np(V) under the initial conditions and their concentrations were 1x10^?4 and 1x10^?3 mol·dm^?3, respectively. The experiments were carried out under the various conditions changing the irradiation intensities of the Hg lamp in the various concentrations of HNO₃. It was found that the rates of the redox reactions of the Pu ions were significantly affected by the irradiated light as well as the acid strength. Under the irradiation of the 0.015 W Hg lamp in 3 M HNO₃ solution containing a tenfold excess of a hydroxylamine and hydrazine, more than 95% Pu(ID) was oxidized rapidly to Pu(IV) within 10 min irradiation and it remained at the same valency even after the continuous further irradiation.

On the other hand, the irradiation did not change the valency of Np(V) under the conditions studied. These valency conditions, i.e. Pu(IV) and Np(V), are appropriate for separating Np from Pu by the solvent extraction with TBP-n-dodecane.

The present results lead to the conclusion that the photochemical method has a high potential for removing Np from the mixed solution of Pu and Np. The photochemical redox reaction mechanisms of Pu and Np in the nitric acid solution were discussed from the stand-points of the thermodynamic and kinetic considerations related to the variation in their standard electrode potentials of the photo-excited ion species by the light irradiation.     

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Low-Temperature Partitioning of Plutonium with High U(VI)-Loading for Fuel Reprocessing

Measurement of the distribution ratios of Pu(IV), U(VI) and HNO₃ at low temperatures and its treatment with DIST code revealed that a high U (VI)-loading of 30% TBP in n-dodecane splits Pu(IV) down to the aqueous phase more strongly than do at 25°C. Based on these findings, flowsheet conditions to separate Pu(IV) from U(VI) were investigated with EXTRA.M code including the distribution equations obtained above. And tentative flowsheets for non-reductive Pu-splitting process at a temperature of 5°C were proposed for fuel reprocessing mainly based on the effects of U (VI)-loading in the solvent and temperature on distribution ratios of Pu(IV) and U(VI). Distribution ratios of the fission products, Zr, Nb, Ru and Ce were also measured to assess their decontamination from U or Pu products in the above process. Finally behavior of Np, in the proposed partitioning process was discussed by analysis with EXTRA. M code and a redox reaction model.     

Solubility of Tributylphosphate in Plutonium Nitrate Solution and Highly Radioactive Liquid Waste Solution

Abstract

The solubility of tri-n-butylphosphate (TBP) in aqueous solutions of plutonium nitrate (PuN) and in highly radioactive liquid waste (HRLW) of PUREX nuclear fuel reprocessing was investigated. By an empirical formula the solubility of TBP in PuN solutions was described in the range of 0–0. 1 M Pu and 1–8M HNO₃ concentrations. The following items were elucidated:

(1) The logarithm of TBP solubility (S) in the solution of interest varies inversely in proportion to the concentration of Pu(IV) in the range of 0–0.1M PU(IV) at a constant concentration of HNO₃, indicating that Pu(IV) simply behaves as an electrolyte for the salting-out of TBP. Log S subsequently levels off with increasing Pu concentration, which would be due to a change in the principal dissolution form of TBP having an interaction with Pu (IV).

(2) The variation in S in PuN solutions (0–0.1M PU) with nitric acid concentration shows almost the same tendency as that in HNO₃ solution.

(3) A dependency of S on fission product metal ions in HNO₃ for HRLW similar to that for PuN was observed.

(4) The logarithm of the ratio of TBP solubility in water to that in solution of interest was nearly proportional to l/T for HRLW solution or for low concentration of PuN solution. That deviates from the linear relation at high temperature when the concentration of PuN is increased, which can be explained by the change in ionic form of Pu.     

Prospective Power-Reactor Fuel Cycles Based on Water-Free Reprocessing of Spent Fuel

A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO₂, PuO₂ or (U, Pu)O₂ from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O₂ fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization.     

Water boiling on the corium melt surface under VVER severe accident conditions

Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the ‘Rasplav-2’ experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO_2+x–16% ZrO₂–15% Fe₂O₃–6% Cr₂O₃–3% Ni₂O₃. The melt surface temperature ranged within 1920–1970 K.     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

Separation of Plutonium-238 from Fission Products by Solvent Extraction Using HDEHP

A procedure for separating ²³⁸Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO₃ solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 10² for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5 mg of ²³⁸Pu was obtained with an efficiency exceeding 95%.     

Ageing of a phosphate ceramic used to immobilize chloride contaminated actinide waste

A process for the immobilization of intermediate level waste containing a significant quantity of chloride using Ca₃(PO₄)₂ as the host material has been developed. Waste ions are incorporated into two phosphate-based phases, chlorapatite [Ca₅(PO₄)₃Cl] and spodiosite [Ca₂(PO₄)Cl]. Non-active trials performed using Sm as the actinide surrogate demonstrated the durability of these phases in aqueous solution. Trials of the process, in which actinide-doped materials were used, were performed at PNNL which confirmed the wasteform resistant to aqueous leaching. Initial leach trials conducted on ²³⁹Pu/²⁴¹Am loaded ceramic at 313 K/28 days gave normalized mass losses of 1.2 × 10⁻⁵ g m⁻² and 2.7 × 10⁻³ g m⁻² for Pu and Cl, respectively. In order to assess the response of the phases to radiation-induced damage, accelerated ageing trials were performed on samples in which the ²³⁹Pu was replaced with ²³⁸Pu. No changes to the crystalline structure of the waste were detected in the XRD spectra after the samples had experienced an α radiation fluence of 4 × 10¹⁸ g⁻¹. Leach trials showed that there was an increase in the P and Ca release rates but no change in the Pu release rate.     

铀钚萃取洗涤-共反萃工艺Ⅰ.串级工艺优化

快堆燃料后处理是实现快堆燃料闭式循环的关键环节之一,快堆乏燃料中裂变产物含量高,进行后处理需要多个铀钚萃取洗涤-共反萃循环才能达到去污效果。本研究针对快堆乏燃料高钚浓度和需要多个萃取洗涤 共反萃循环净化裂变产物的特点,采用模拟料液通过多次串级实验,确定了满足铀钚收率及避免钚聚合的铀钚萃取洗涤-共反萃工艺,实验结果表明,1A铀、钚萃取收率分别为99.995%和99.996%,1B铀、钚反萃收率分别为99.936%和99.996%。     

在线进样ICP-MS用于239Pu气溶胶活度浓度连续监测技术研究

为了弥补传统场所α气溶胶实时连续监测技术在后处理等场所239Pu气溶胶活度浓度连续监测方面探测能力和响应速度方面的不足,本文开展了基于ICP MS的空气中239Pu气溶胶连续监测技术研究。通过对气溶胶直接进样方法、ICP MS连续测量定量方法、238U干扰等方面研究工作的开展,建立了基于ICP MS、气溶胶直接进样系统和膜去溶雾化器联用的场所239Pu气溶胶活度浓度连续监测技术。在单次测量时间为1 min的情况下,依据ISO 11929（2019）计算得到本系统对239Pu气溶胶判断限和探测限分别为124×10-5 Bq/m3和248×10-5 Bq/m3,优于传统的基于PIPS的α气溶胶实时连续监测系统,远低于239Pu的导出空气浓度限值,证明该技术可为相关工艺场所239Pu气溶胶活度浓度连续监测提供快速高效的技术支持。     

Effect of a Boron Screen on the Sensitivity of Neutron Fission Chambers

It is suggested that absorbing screens with ¹⁰B be used to maintain constant sensitivity under prolonged irradiation of fission chambers with natural uranium. The transmission factor T (E) of boron screens with various thicknesses ( = 0.1–2 ge/cm²) for a wide neutron energy range and attenuation of a spectrum of the type _e/E are estimated. The group and average group constants of the transmission factor of boron are calculated for neutron fluxes in 25 energy groups of the neutron cross sections library.The contribution of ²³⁸U and ²³⁵U to the signal of a fission chamber with natural uranium is analyzed as a function of the boron screen thickness. ²³⁹Pu accumulation and ²³⁸U burnup are estimated using ²³⁸U group capture cross sections, ²³⁸U and ²³⁹Pu fission cross sections, and the group values T (E)E/E obtained by the authors. It is shown that in the absence of a boron screen for thermal-neutron fluence 10¹⁷ cm^–2 the sensitivity of a fission chamber with natural uranium increases as a result of the formation of ²³⁹Pu. A boron screen with = 1 g/cm² makes it possible to maintain the sensitivity of the fission chamber constant up to thermal-neutron fluence 5·10²² cm^–2.     

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO₃ concentrations. The flowsheet with a supply of 0.15 mol/dm³ HNO₃ solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×10⁵ and 2.42×10⁵, respectively.     

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.     

Neutron absorber coatings for criticality control in annular tanks employed in spent fuel reprocessing

Computational analysis has been carried out to evaluate the effectiveness of neutron absorber coatings for criticality control in an annular tank used in fast reactor spent fuel reprocessing unit. The effect of composition, thickness and coating configuration for a given tank design and fuel solution concentration was evaluated on the basis of the multiplication factor (k_eff) calculated using the Monte Carlo N-Particle (MCNP) code. The neutron absorbers considered for the study were pure boron carbide (B₄C), B₄C/Ni–Cr combination and colmonoy. The effect of enriched boron was also analyzed. The results show that the coatings can enhance the storage capacity up to 30% for the annular tank studied.     

Hypothetical integration of a fast reactor to the fuel cycle in Slovakia

The inspiration for dealing with the topic of fuel cycle back-end was attendance at a European project called RED-IMPACT – Impact of Partitioning Transmutation and Waste Reduction Technologies. This paper includes an image how to re-use energetic potential of stored spent fuel and at the same time how to effectively reduce spent fuel and radioactive waste volumes aimed for deep repositories. The first part is based on the analysis of Pu and minor actinides (MA) content in actual VVER-440 spent fuel stored in Slovakia. The next parts present the hypothetical possibilities of reprocessing and Pu re-use in a fast reactor under Slovak conditions. For the hypothetical transmutation of heavy nuclides (Pu and MA) contained in Slovak spent fuel a SUPERPHENIX (SPX) fast reactor with increased power was chosen because a fast nuclear reactor cooled by sodium belongs to the group of Generation IV reactor systems. This article deals with the analysis of power production and fuel cycle indicators. The indicators of the SPX calculation model were compared with the results of the VVER-440 spent fuel with the initial fuel enrichment of 4.25% U-235 + 3.35% Gd₂O₃. The created SPX model in the spectral computer code HELIOS 1.10 consists of a fissile (fuel) and a fertile part (blanket). All kinds of calculations were performed by the computer code HELIOS 1.10. This study also exposes the HELIOS modelling and simulating borders.     

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

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

Supposed existence of Np4+ in a genuine dissolver solution from the results of extraction simulation by PARC-L code

With a genuine spent fuel solution (a dissolver solution), a laboratory-scale reprocessing experiment of an extraction–separation process was performed using mixer-settlers as extractors. In the experiment, n-butyraldehyde was utilized as a reducing reagent of Np^(VI)O₂²⁺ to Np^(V)O₂⁺ for the purpose to distinguish Np^(VI)O₂²⁺ from Np⁴⁺. From the Np concentration in the aqueous phase, Np would be extracted from the dissolver solution together with U and Pu. The scrutiny of Np behavior was performed utilizing 66 cases of calculation results by a Japan Atomic Energy Agency open extraction simulation code, the Program for Advanced Extraction with Radiation Effect Calculation–Lightened version. From the scrutiny, the authors found that the calculation result with 60% of Np⁴⁺ in the dissolver solution represented the best experimental extraction–separation behavior of Np. Therefore, it was supposed that the dissolver solution contained sufficient proportion of Np⁴⁺ to affect the extraction–separation behavior of Np.