Production Of 238Pu Research Articles

High-resolution neutronics model for 238Pu production in high-flux reactors

High-resolution neutronics model for 238Pu production in high-flux reactors

Assessment of the possibility for large-scale 238Pu production in a VVER-1000 power reactor

The paper presents the estimates for the possibility for large-scale production of 238Pu in the core of a VVER-1000 power reactor. The Np-fraction of minor actinides extracted from transuranic radioactive waste is proposed to be used as the starting material. The irradiation device with NpO2 fuel elements is installed at the reactor core center. The NpO2 fuel lattice pitch is varied and the irradiation device is surrounded by a heavy moderator layer to create the best possible spectral conditions for large-scale production (~ 3 kg/year) of conditioned plutonium with the required isotopic composition (not less than 85% of 238Pu and not more than 2 ppm of 236Pu). Plutonium with such isotopic composition can be used as the thermal source in thermoelectric radioisotope generators and in cardiac pacemakers. It has been demonstrated that the estimated scale of the 238Pu production in a VVER-type power reactor exceeds considerably the existing scale of its production in research reactors.

Application of Minor Actinides for the Large-Scale Production of 238Pu in a Supercritical Light-Water Reactor

Application of Minor Actinides for the Large-Scale Production of 238Pu in a Supercritical Light-Water Reactor

238Pu production: a State-of-the-Art review of NpO2 target fabrication technologies and processing of irradiated targets

Abstract Space mission beyond the solar system cannot rely on photovoltaic (PV) cells as primary power source, and combinations of PV cells and batteries. For such purpose, Radioisotope Power Systems powered with 238Pu have served well for all flagship space exploration missions since the early Apollo missions until today’s Mars Exploration Program and continue to be the preferred primary energy source for future missions. 238Pu production proceeds via neutron irradiation of 237Np, which is created as a by-product in nuclear fission reactors. Neutron irradiation of 237Np to produce 238Pu is conceptually very simple, but the production of sizeable quantities of 238Pu with acceptable isotopic purity, and the further separation and processing stages pose formidable technological challenges. 238Pu is a highly radiotoxic alpha emitter with an elevated specific activity and high decay energy with high risk for the workers, and its precursor, 237Np, is a radiotoxic alpha emitter whose daughter 233Pa decays to 233U with a strong gamma emission. Furthermore, the very rich and weakly explored chemistry of Np is also a challenge. The historical 238Pu inventory of US-DOE has been highly consumed and supplies of available 238Pu to support new missions have diminished. NASA and US-DOE have started an extensive project to re-establish 238Pu production for US space missions. In Europe, ESA has also shown some interest in an European production of 238Pu for the European space missions. In the present review, the manufacture of Np targets for 238Pu production by irradiation, and the target processing are discussed and assessed from an European production perspective by comparing mainly US state-of-the-art with the European know-how and the current facilities. Two principal options for target fabrication stand out: aluminium-clad NpO2–Al CERMET and zircaloy-clad full-ceramic NpO2 targets. The principal advantages of the well documented CERMET route can be found in the irradiation stage. These targets allow a high flexibility in heavy metal loading fraction and have high thermal performance. But, they have significant drawbacks in terms of the amount and type of nuclear waste that is generated at the back-end. Instead, the full-ceramic targets are conceptually very similar to conventional nuclear fuels, and would be preferred from back-end perspective, although a demonstration of large scale production remains yet to be done. Presently, there are no operational large scale Np target production, irradiation, processing or 238Pu handling facilities in Europe, but the nuclear infrastructure exists and is operational to study each of the steps at least at small scale and all steps have similarities with well-established industrial capabilities in Europe.

Effects of multi-recycling of uranium reprocessed from spent nuclear fuel on the proliferation resistance of plutonium and uranium

The effects of re-enriching and multi-recycling reprocessed uranium (RepU) from the spent nuclear fuel of light water reactors on the proliferation resistance (PR) of Pu and U are studied. The presence of 236U and 238Pu enhances the PR of multi-recycled RepU. The 236U/235U ratio of RepU varies from 0.4 to 1.2, depending on the initial 235U enrichment and fuel burnup. 236U is co-enriched while re-enriching 235U in RepU. Based on the centrifuge enrichment data, the enrichment coefficient of 236U (α236), defined as 236U/235U ratio after and before enrichment, is 0.703, on average. The α236 value is utilized in a simple model to estimate the attainable maximum possible 235U enrichment for multi-recycled RepU. Estimations showed an increase in the PR of multi-recycled and re-enriched RepU, which was confirmed by matched-abundance ratio cascade model. The PR of Pu is enhanced in each recycle due to the increased production of 238Pu from 236U.

Conceptual design study on Plutonium-238 production in a multi-purpose high flux reactor

Plutonium-238 has always been considered as the one of the promising radioisotopes for space nuclear power supply, which has long half-life, low radiation protection level, high power density, and stable fuel form at high temperatures. The industrial-scale production of 238Pu mainly depends on irradiating solid 237NpO2 target in high flux reactors, however the production process faces problems such as large fission loss and high requirements for product quality control. In this paper, a conceptual design study of producing 238Pu in a multi-purpose high flux reactor was evaluated and analyzed, which includes a sensitivity analysis on 238Pu production and a further study on the irradiation scheme. It demonstrated that the target structure and its location in the reactor, as well as the operation scheme has an impact on 238Pu amount and product quality. Furthermore, the production efficiency could be improved by optimizing target material concentration, target locations in the core and reflector. This work provides technical support for irradiation production of 238Pu in high flux reactors.

Application of Np–Am Mixture in Production of 238Pu in a VVER-1000 Reactor and the Reactivity Effect Caused by Loss-of-Coolant Accident in the Central Np–Am Fuel Assembly

This paper presents the results obtained from numerical evaluations for the possibility of large-scale 238Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of 238Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor.

Neptunium mononitride as a target material for Pu-238 production

Deep space exploration requires specialized sources for both thermal and power applications. Radioactive decay heat of plutonium-238 (238Pu) provides these sources in the form of radioisotope thermoelectric generators (RTGs). The238Pu is producedvianeutron capture reaction involving neptunium-237 (237Np) target material. Continual optimization of237Np target materials and evaluation of potential alternative targets for production of238Pu RTGs are advantageous for meeting ongoing space power system resource requirements. Current production of238Pu for RTGs for the United States space program utilizes neptunium dioxide (237NpO2) targets; however, the use of neptunium mononitride (237NpN) presents an opportunity to increase the mass of237Np per target compared to the dioxide form, as well as increase the thermal conductivity of the target. To assess the viability of a237NpN target material, the material chemistry must be thoroughly evaluated, including synthesis methods and dissolution and reprocessing schemes. This review presents a summary of synthesis pathways for237NpN based on published literature on actinide mononitrides. Specific literature on237NpN is limited, necessitating evaluation of other actinide systems to gather parallels. This suggests a need for additional experimental studies on237NpN. A particular limitation in the existing literature is a lack of information on the differences in material characteristics, such as morphology, particle size, and trace chemical impurities, as a function of synthesis method. These parameters may affect subsequent reactor performance or dissolution of irradiated targets. The evaluation of existing literature is presented with a focus on the efficacy of237NpN targets for238Pu production.

A Possibility for Large-Scale Production of 238Pu in Light-Water Reactor VVER-1000

This paper considers the possibility for large-scale production of plutonium isotope 238Pu in the light-water nuclear power reactor VVER-1000. 238Pu is a unique source of long-term autonomous energy supply in various devices for remote regions of the Earth and in outer space. The design of the irradiation device with 237NpO2 as a starting material is proposed, which is placed in the central zone of the VVER-1000 reactor core and makes it possible to achieve 8% of the specific Pu production (Pu/237Np) by optimizing the pitch of NpO2-rod lattice. The computations showed that the scale of 238Pu production in the irradiation device was remarkably larger (2 ÷ 7 times more) than similar values in research reactors. At the same time, the use of heavy neutron moderators with low neutron absorption (natural lead or lead isotope 208Pb) around the NpO2 fuel assembly (FA) made it possible to obtain high-purity 238Pu with the content of 236Pu below 2 ppm. The paper also shows that if the irradiation device is placed in central zone of the VVER-1000 reactor core, then the displacement damage dose in the reactor vessel remains low enough to conserve its strength properties throughout the entire period of the reactor operation (60 years).

Recent 238Pu Production Activities at Idaho National Laboratory

The Plutonium-238 (238Pu) Production program at Idaho National Laboratory (INL) is actively qualifying irradiation targets containing 237Np for the Advanced Test Reactor (ATR) to produce 238Pu for future National Aeronautics and Space Administration missions. INL qualified and loaded seven targets in the ATR’s south flux trap for cycle 169A, which occurred in Spring 2021. The irradiation qualification program has expanded to additional ATR irradiation positions after two baseline production targets in three positions validated significant production of 238Pu. The validation model was followed by the PFS-1 experimental test in the ATR Critical Facility that verified 238Pu production cross sections. This paper outlines the progress and status of the 238Pu production program at INL. The qualification effort, safety analysis, hardware status, and future activities for qualification of an updated target design for use in the ATR are discussed.

Measurement of the Effective Capture Cross Section of 238Np in the High Flux Isotope Reactor

The cross sections of 237Np and 238Np are important for accurate modeling and simulation of 238Pu in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Uncertainties in these cross sections can impact the ability to predict and optimize the target design and loading for 238Pu production targets. The effective capture cross section of 237Np in the location of pneumatic tube 1 in HFIR was measured as a first step in the measurement of the 238Np capture and fission cross sections. This paper describes the flux measurements, 237Np experiments, and data analysis of the 237Np capture cross section in HFIR.

Plutonium-238 Production Program Results, Implications, and Projections from Irradiation and Examination of Initial NpO2 Test Targets for Improved Production

An alternative target design with potential improvements, including a major increase in 238Pu production rate and annual capacity; fewer targets to be fabricated, irradiated, and processed; and a significant replacement of a large volume of caustic-nitrate, aluminum-bearing radioactive liquid waste with a smaller volume of solid metal waste, has been conceived and evaluated using reactor physics and thermal-hydraulic analyses. The alternative target design uses pressed pellets of 237NpO2, sintered to 92% to 93% of theoretical density, and stacked inside a Zircaloy-4 cladding tube. Four test targets were fabricated, irradiated, and examined. No melting or other potential problems were indicated. Projections from measured constituents indicated annual production could be increased by a factor of ~2, and the number of targets required to be fabricated, irradiated, and processed could be reduced by a factor of ~5.

An Alternative Solvent Extraction Flowsheet for Separating 237Np from 238Pu for Space Power Applications

ABSTRACT The production of 238Pu from 237Np, to the specifications required for space flight, requires the separation of plutonium from neptunium and a range of fission products. The valuable neptunium is then recycled into targets for further irradiation. At Oak Ridge National Laboratory (ORNL), a solvent extraction process has been developed to separate neptunium and plutonium that uses sodium nitrite to control the neptunium oxidation. This results in a neptunium product contaminated with sodium, which needs to be removed prior to recycling. We have developed a separation process that results in a sodium-free neptunium product with very low levels of plutonium contamination, there by reducing the number of purification steps and thus simplifying the process.

Combination of DGA and LN Columns: A Versatile Option for Isotope Production and Purification at Oak Ridge National Laboratory

ABSTRACT The chromatographic resins DGA and LN have been used sequentially to separate curium, californium, plutonium, and various fission products (FPs) to obtain products with a pure radioisotope or a radioelement with high isotopic purity. This 2-step combination of resins has found applications in the 252Cf and 238Pu campaigns performed at Oak Ridge National Laboratory. Experiments have been carried out in glove boxes, shielded caves, and hot cells, demonstrating that these resins are adequate and maintain their performance in highly radiolytic environments. Using these resins has led to the recovery of curium highly enriched (>95%) in 248Cm, the clean separation of 251Cf for the superheavy element research, and the recovery of 147Pm from the FP stream obtained during the 238Pu production campaign while demonstrating that the raffinate post DGA is essentially free of transuranic elements.

Analysis of producing 238Pu as a byproduct in an MSFR

238Pu can be used as the heat source in radioisotope thermoelectric generators for deep space exploration, while it is currently in shortage because of the suspension of 238Pu production in the USA and Russia for many years. Two possible scenarios for the production of 238Pu as a byproduct in a molten salt fast reactor (MSFR) are assessed in this work. For the first scenario, the MSFR is started with low-enriched uranium (LEU), and neptunium is extracted from the fuel salt and then fed into the fertile salt of thorium for high-purity 238Pu production. For the second scenario, the ignition material for the MSFR is 233U, and highly purified 238Pu is produced by directly extracting plutonium from the fuel salt. The simulated results indicate that on average 15.48 kg 238Pu with the quality (ratio of 238Pu mass to total mass of Pu isotopes) larger than 98.6% can be produced per year during a 60-year operation for scenario I. In parallel, 8.31 kg 238Pu with the quality larger than 91.9% can be produced annually for scenario II. The 238Pu production has a small impact on the neutronic performance of the core for the two scenarios except that bred 238U is reduced by 922 kg for scenario I and by 261 kg for scenario II during the 60-year lifetime. The 238Pu production has a very tiny effect on the radiotoxicity of discharged nuclear fuel for scenario I, while it can reduce the total radiotoxicity of discharged nuclear fuel by three times for scenario II.

Using N,N-dialkylamides for neptunium purification from other actinides for space applications

ABSTRACT Tributylphosphate (TBP) has been reliably used in the solvent extraction of actinides for decades. One current application of TBP is the purification of plutonium-238 (238Pu), a heat source for radioisotope thermoelectric generators. Part of the 238Pu production process includes irradiating neptunium-237 (237Np) previously purified using di-(2-ethylhexyl)phosphoric acid (HDEHP). However, radiolytic decomposition of TBP and HDEHP forms several problematic phosphate-based degradation products. These non-incinerable byproducts contaminate the final 238Pu product and complicate the recycle of 237Np. In contrast, alternative extractants, such as the N,N-diakylamides, have incinerable byproducts. Nonbranched N,N-dialkylamides (N,N-dihexylhexanamide (DHHA) and N,N-dihexyloctanamide (DHOA)) and branched N,N-dialkylamides (N,N-di(2-ethylhexyl)butanamide (DEHBA) and N,N-di(2-ethylhexyl)isobutanamide (DEHiBA)) were considered for Np/Pu and Np/Th separations to replace TBP or HDEHP in 238Pu production. To understand the extraction potential of N,N-dialkylamides, all four N,N-dialkylamides were used to extract actinides (Th, U, Np, Pu, Am) across a range of solution-accessible oxidation states. Those results were then compared based on the actinides’ oxidation state (III - VI) to illustrate some of the similarities and differences in extraction behavior.

Design Studies for the Optimization of 238Pu Production in NpO2 Targets Irradiated at the High Flux Isotope Reactor

Efforts to reestablish a domestic 238Pu production capability in support of National Aeronautics and Space Administration mission objectives are ongoing throughout the U.S. Department of Energy complex. The Plutonium-238 Supply Project (PSP) was initiated in response to a report published by the National Research Council in 2011 stating that “without a restart of 238Pu production, it will be impossible for the United States, or any other country, to conduct certain important types of planetary missions after this decade.” The PSP is targeting a sustained, constant production rate of 1.5 kg/year of heat source PuO2 for several years. Design and optimization studies of 237Np-bearing targets are underway at Oak Ridge National Laboratory (ORNL). It is anticipated that targets will be irradiated in ORNL’s High Flux Isotope Reactor (HFIR) and in the Advanced Test Reactor (ATR) at Idaho National Laboratory. A variety of target materials, containments, arrangements, and irradiation histories have been analyzed, and the results indicate that a sufficient quantity of 238Pu can be produced in HFIR and ATR to fulfill the PSP’s constant production rate target. This paper focuses on the design and optimization of new target configurations containing pellets that are (1) ~93% of the theoretical density of NpO2, (2) loaded into pins of cladding materials that can be handled as solid waste following postirradiation 238Pu recovery operations, (3) irradiated in various vertical experiment facility (VXF) locations in the HFIR permanent beryllium reflector, and (4) rotated within and/or moved to another VXF location following each HFIR operational cycle to maximize 238Pu production and minimize peak heat generation rates.

Thermomechanical Safety Analyses for a 238Pu Production Target at the HFIR

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify irradiation of production targets containing neptunium dioxide/aluminum cermet (NpO2/Al) pellets for the production of plutonium-238 (238Pu). High heat generation rates (HGRs) due to a fertile starting material (237Np), low melting temperatures, and previously unstudied material irradiation behavior (i.e., swelling/densification, fission gas release) require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. Experience gained from previous models for preliminary target designs is incorporated into a more comprehensive production target model designed to qualify a target for three cycles of irradiation and illuminate potential in-reactor behavior of the target.

The Separation of 241Am from Aged Plutonium Dioxide for Use in Radioisotope Power Systems Using the AMPPEX Process

Electrical power sources used in outer planet missions are a key enabling technology for data acquisition and communications. State–of-the-art power sources generate electricity from alpha decay of 238Pu via thermoelectric conversion. However, production of 238Pu requires specialist facilities including a nuclear reactor, a source of 237Np for target irradiation and hotcells to chemically separate neptunium and plutonium within the irradiated targets. These specialist facilities are expensive to build and operate, so naturally, a more economical alternative is attractive to the space industry. Within Europe 241Am is considered a promising alternative heat source for radioisotope thermoelectric generators (RTGs) and radioisotope heating units (RHUs) since, as a daughter product of 241Pu decay, 241Am exists in 1000 kgs quantities within the UK civil plutonium stockpile. A chemical separation process is required to extract the 241Am in a pure form and this paper describes the AMPPEX process (Americium and Plutonium Purification by Extraction), successfully developed over the past five years to isolate 241Am in high yield (> 99%) and to a high purity (> 99%). The process starts by dissolving plutonium dioxide in nitric acid with the aid of a silver(II) catalyst, which is generated electrochemically. The solution is then conditioned and fed to a PUREX type solvent extraction process, where the plutonium is separated from the americium and silver. The plutonium is converted back to plutonium dioxide and the americium is fed forward to a second solvent extraction step. Here the americium is selectively extracted leaving the silver in the aqueous phase. The americium is stripped from the solvent and recovered from solution as americium oxalate, which is calcined to give americium dioxide as the final product. This paper describes the development of the AMPPEX process over a series of six solvent extraction separation trials using centrifugal contactors. The technical underpinning reported here has allowed the design of a facility capable of producing 25-50 g of americium per day.

Isotopic correlation for 242Pu composition prediction: Multivariate regresssion approach

Multivariate regression calibration using multiple linear regression (MLR), principle component regression (PCR) and partial least squares regression (PLSR) algorithm was performed on 238Pu, 239Pu, 240Pu and 241Pu atom% abundances to predict 242Pu isotopic abundance. The MLR algorithm was found to be the best among these three algorithms. The effect of 238Pu composition on the 242Pu abundance prediction was found to be small but significant especially for achieving high accuracy of <0.5%. PCR and PLSR generated nearly identical results and were inferior to the MLR results. A comparison of MLR results with those obtained by employing seven previously reported empirical methods revealed far superior prediction capability of MLR model. Among the seven empirical models, the best prediction capability was found for Bignan correlation containing 238Pu isotopic data. The study clearly demonstrates that the production of 238Pu and 242Pu has some small correlation and the use of 238Pu in isotopic correlation for 242Pu prediction is important to get accurate results.