Hot Cell Facility Research Articles

Westinghouse Churchill Hot Cell Facility and Capabilities

Westinghouse Churchill Hot Cell Facility and Capabilities

Augmentation of hot cell facility for pyro-process of irradiated metal fuels

Hot cells with an inert atmosphere are used for handling air-sensitive, hygroscopic, and radioactive materials like mixed carbide fuels and for pyro-processing metallic fuels. Pyro-processing of spent metal fuels was based on the electrochemical recovery of actinides in high-temperature molten salts (LiCl-KCl eutectic with 58.5 mol.% LiCl). To demonstrate the remote operation feasibility of the electro-refining process of irradiated U-6 wt% Zr inside hot cells and also to study the process parameters, a pilot scale (∼100 g) pyro-process facility was set up in one of our hot cells. For this purpose, the existing hot cell facility was augmented with necessary modifications and customized to the experimental needs. This paper discusses in detail the augmentation works carried out on the existing facility to facilitate the remote operation of electro-refining of irradiated metallic fuels.

Radiochemical and chemical characterization of fuel, salt, and deposit from the electrorefining of irradiated U-6 wt% Zr in hot cells

Abstract Metal fuels are considered as the promising candidates for future fast breeder reactors. Pyro-chemical reprocessing is the ideal method for reprocessing spent metallic fuels due to the inherent process advantages. Electrorefining run was demonstrated in a hot cell facility with irradiated U-6 wt% Zr alloy at 500 °C using LiCl–KCl eutectic melt. In order to understand the behavior of the actinides and various fission products during high-temperature electrolysis, various process streams, viz., irradiated metal alloy fuel, the eutectic salt, and the cathode deposit were analyzed for the uranium, plutonium, and other fission product contents. Various methods employed for characterizing the process streams and the behaviors of some of the fission products during the electrolysis process are highlighted. The major gamma emitting radionuclides present in the irradiated fuel were 106Ru, 125Sb, 134Cs, 137Cs, 144Ce, and 154Eu. During electrorefining, cesium, cerium and europium were oxidized and dissolved in the molten media, whereas ruthenium and antimony remained in the anode basket. A minor contamination of zirconium was found in the cathode deposit.

Actinide targets for the synthesis of superheavy nuclei

The use of heavy actinide targets, including 243Am, 240,242,244Pu, 245,248Cm, 249Bk, and 249Cf, irradiated by intense heavy ion beams of 48Ca has resulted in a significant expansion of the periodic table since 2000, including the discovery of five new heaviest elements and more than 50 new isotopes. These actinide materials can only be produced by intense neutron irradiation in very high flux reactors followed by chemical processing and purification in specialized hot cell facilities available in only a few locations worldwide. This paper reviews the reactor production of heavy actinides, the recovery and chemical separation of actinide materials, and the preparation of actinide targets for superheavy element experiments. The focus is on 248Cm, 249Bk, mixed 249−251Cf, and 254Es, including current availabilities and new production processes. The impacts of new facilities, including the Superheavy Element Factory at Dubna, accelerator and separator upgrades at RIKEN, and proposed upgrades to the High Flux Isotope Reactor at Oak Ridge are also described. Examples of recent superheavy element research are discussed as well as future opportunities for superheavy research using actinide targets.

Contributions of MYRRHA to the European Fusion Energy Roadmap

One of the main objectives of the EUROfusion Roadmap to Fusion Energy during Horizon Europe is to finalize the conceptual design (thus preparing the engineering design phase) and initiate the construction of a fusion spectrum neutron source (IFMIF-DONES). The Belgian Nuclear Research Center is currently developing the MYRRHA research facilities, which include a subcritical accelerator-driven system ignited by a 600 MeV superconducting linear proton accelerator coupled to a 100 MeV proton injector. The implementation of the multipurpose fast spectrum reactor MYRRHA will proceed in phases, with phase I, covering 2019–2027, including the deployment of the 100 MeV proton accelerator and associated fusion material research facilities. The research on fusion materials on the basis of the 100 MeV proton beam and with hot cell facilities will focus on innovative materials and composites and will be fully complementary to IFMIF-DONES as well to the planned research program on using currently available nuclearized facilities such as BR2 (Belgian Research Reactor 2) and Hot Cells for mechanical, microstructural and chemical analysis. A bespoke fusion irradiation module is also foreseen on the MYRRHA reactor, the construction of which is planned in the third phase of the implementation of the project (i.e. after 2035). Provisions have been made in the core of the MYRRHA reactor to include so-called In-Pile Sections around the central core or at the central core position below the spallation target when operating in subcritical mode. In this contribution, we review the role and technical capabilities that MYRRHA facilities may provide in the framework of the European Fusion Electricity Roadmap and specifically for DEMO Engineering Design, Material's Development and enhancement of Technology Readiness Level. Aspects of education and training, being an integral part of the European nuclear fusion program, are also touched upon. Finally, the synchronization of the timelines for the implementation of MYRRHA and the EU fusion facilities are discussed.

The Radiation Impact of Delayed Photoneutron Production in ITER Components

Photoneutrons may be generated in beryllium by energetic gamma rays via the reaction 9Be(γ,n)8Be. In ITER, the beryllium layer of the first wall may be the source of such photoneutrons. During plasma operation, these are of insignificant intensity compared with D-T neutrons from the plasma, but after shutdown, photoneutrons produced by decay gammas from neutron-activated material may be significant enough to impact sensitive electronic components in diagnostic or remote handling equipment that would not otherwise be exposed to neutrons. Studies have been performed to characterize the expected photoneutron source and to evaluate the fluxes arising in detailed three-dimensional models of the ITER tokamak. The results show photoneutron fluxes approaching 105 n/cm2·s within the vessel and up to 103 n/cm2·s elsewhere within the bioshield 14 days after shutdown. When first-wall panels are being transported to the Hot Cell Facility after irradiation, a photoneutron flux exceeding 104 n/cm2·s within the transfer cask is predicted 21 days after shutdown. The peak values in the surrounding building are between 102 and 103 n/cm2·s at the same time.

Electrorefining of un-irradiated and irradiated U-6 wt. % Zr metal fuel for deposition of uranium

The present paper reports the data on electrorefining of un-irradiated and irradiated U-6 wt. % Zr (U-Zr) fuel carried out at 773 K in LiCl-KCl-UCl3 electrolyte in an augmented hot cell facility. The electrorefining cell and electrode handling system were specially designed, fabricated, and installed inside the hot cell. Potentiostatic electrorefining was employed for the selective deposition of uranium. About 61 % and 36 % of initial uranium feed was deposited during the electrorefining of un-irradiated and irradiated U-Zr fuel respectively. Current efficiencies of about 96 % and 80 % were obtained for un-irradiated and irradiated runs respectively. No significant change in concentration of uranium in the electrolyte was observed before and after the electrorefining. The co-deposition of other fission products was negligible and zirconium content in the uranium product was found to be 3 mg/g of uranium.

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. To accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300∘C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8×1021n/cm2 fast fluence (En>1.0MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

Fusion-dedicated material irradiation facilities at MYRRHA: Conceptual design and damage equivalence studies

The Belgian Nuclear Research Centre is currently developing MYRRHA - a subcritical accelerator driven system ignited by a 600 MeV superconducting linear proton accelerator, cooled with an eutectic lead bismuth. The implementation of MYRRHA will take place in phases, where phase I, covering 2019–2027, includes the deployment of the 100 MeV proton accelerator and beam-added research facilities. Development of the dedicated facilities to enable irradiation of fusion materials and components is in the portfolio of this R&D block. The baseline concept of the 100 MeV fusion-dedicated facilities has been presented for the first time in 2019 and in this contribution, we extend its description and provide main elements of the conceptual design of the irradiation installations, preliminary configuration of the hot cell facilities (required to load/unload the irradiation modules), information on the irradiation envelope. In addition, we summarize the results of the computational study performed to investigate the equivalence (in terms of damage rate, nuclear transmutation, temperatures and expected irradiation-induced microstructure) of the irradiation conditions at MYRRHA facilities in comparison with other material test facilities such as IFMIF and ESS, and in DEMO as well as in the mixed-spectrum material test reactors. The aspects related to miniaturization of the sample geometries are also discussed to substantiate the feasibility and relevance of this facility for fusion material's R&D programme.

An absolute measurement of the neutron production rate of a spent nuclear fuel sample used for depletion code validation

A method to determine the neutron production rate of a spent nuclear fuel segment sample by means of non-destructive assay conducted under standard controlled-area conditions is described and demonstrated. A neutron well counter designed for routine nuclear safeguards applications is applied. The method relies on a transfer procedure that is adapted to the hot cell facilities at the Laboratory for High and Medium level Activity of SCK CEN in Belgium. Experiments with 252Cf(sf) sources, certified for their neutron emission rate, were carried out at the Joint Research Centre to determine the characteristics of the detection device. Measurements of a segment of a spent nuclear fuel rod were carried out at SCK CEN resulting in an absolute and non-destructive measurement of the neutron production rate avoiding any reference to a representative spent nuclear fuel sample to calibrate the device. Results of these measurements were used to study the performance of depletion codes, i.e., ALEPH2, SCALE, and Serpent2. The study includes a code-to-code and code-to-experiment comparison using different nuclear data libraries.

Overview of recent ITER TBM Port Plug R&D activities

The objective of ITER project is the demonstration of the scientific and technological feasibility of fusion energy. Tritium fuel is required for the fusion reaction in DEMOnstration reactors and future commercial reactors but it is not available in nature because of its instability. To investigate tritium breeding self-sufficiency, two equatorial ports in ITER are dedicated to operating test blanket modules (TBMs) using the TBM Port Plug (TBM-PP). The TBM-PP design should be investigated with the physical mock-ups to validate the leak tightness in the primary vacuum boundary and to ensure TBM replacement and refurbishment compatible with Remote Handling (RH) operations in the ITER Hot Cell Facility during the ITER lifetime. This paper summarizes the achievements on the two most relevant R&D topics as described herewith. Several TBM-PP components (i.e., back parts of the TBM-Frame, the Dummy-TBMs, the TBM-Shields, and the TBM feeding pipes) are part of the vacuum boundary implying that these components are Vacuum Quality Classification (VQC) 1A. To test and develop high vacuum sealing for the TBM-Frame flange and the TBM-Set or Dummy-TBM flange, the ITER Large Seal Test Rig (LSTR), and metallic gasket seals for the TBM application have been manufactured and installed in the ITER site. The helium leak tests were performed at different operating temperatures to investigate the vacuum performance of metallic gasket seals with VQC 1A. The TBM-PP design shall guarantee the feasibility of a rapid replacement and refurbishment compatible with ITER Remote Handling (RH) operations by using RH tools in the Hot Cell Facility during the ITER's lifetime. Therefore, an experimental program has been performed to demonstrate the feasibility of the critical RH refurbishment tasks including the insertion and removal of TBM-Set/Dummy-TBM and the RH bolting.

Radiochemical quality control of the radiopharmaceutical, 89SrCl2 produced in FBTR

89SrCl2 radiopharmaceuticals is mainly used for bone pain palliation in the cancer patients, is being produced in FBTR via 89Y(n, p)89Sr using yttria target. The irradiated yttria target is chemically processed in high pure quartz distilled nitric acid medium in hot cell facility, to avoid the corrosion of components of hotcell due to chloride ions while using HCl medium. Being ionic species, the purified 89Sr(II) cation in aqueous solution containing bulk nitrate and other trace anions, exists as SrXn species where X: F−, Cl−, Br−, NO3−, PO43− and SO42−, n: stoichiometric anion content. The aim of the manuscript is to standardise an efficient ultra-low level anion purification method (ppb range) for the conversion of SrXn to SrCl2 and estimate the residual anionic impurities as recommended by the appropriate source specifications for its medical application. Various methods were standardised for the removal of anions in the SrCl2 source produced by the above process which include evaporation, calcination, anion exchange column, cation exchange column as well as its combination with pre-concentration column of ion chromatography (IC) technique using 89Sr tracers as well as FBTR produced 89Sr solution. Assay of 89Sr and other anions including nitrate for the above study were accomplished using Cerenkov counting and ion chromatography respectively. Thus evaporation-calcination-column chromatography mode was finalised to obtain pure SrCl2 source free from nitrate and other anionic impurities. This is the first ever systematic study for the Radiochemical quality control of nca 89SrCl2 radiopharmaceutical produced in a fast reactor. This study also finds its application to any analytical lab as well as industry where there is a requirement of anion purification in the ppb level.

Erratum: Radiation safety evaluation of workers on aerosol generated during radioactive waste processing in hot cell facility

Erratum: Radiation safety evaluation of workers on aerosol generated during radioactive waste processing in hot cell facility

Development of classification criteria for non-reactor nuclear facilities in Korea

Non-reactor nuclear facilities are increasing remarkably in Korea combined with advanced technologies such as life and space engineering, and the diversification of the nuclear industry. However, the absence of a basic classification guideline related to the design of non-reactor nuclear facilities has created confusion whenever related projects are carried out. In this paper, related domestic and international technical guidelines are reviewed to present the classification criteria of non-reactor nuclear facilities in Korea. Based on these criteria, the classification of structures, systems and components (SSCs) for safety controls is presented. Using the presented classification criteria, classification of a hot cell facility, a representative non-reactor nuclear facility, was performed. As a result of the classification, the hot cell facility is classified as the hazard category 3, accordingly, the safety class was classified as non-nuclear safety, the seismic category as non-seismic (RW-IIb), and the quality class as manufacturers’ standards (S).

Radiation safety evaluation of workers on aerosol generated during radioactive waste processing in hot cell facility

Radiation safety evaluation of workers on aerosol generated during radioactive waste processing in hot cell facility

Effects of neutron irradiation and post-irradiation annealing on pop-in crack propagation instabilities in 6061 aluminium alloy

This study analyses the increase of crack propagations instabilities (pop-in) with irradiation during fracture toughness testing of aluminium alloy 6061 alloy after neutron irradiation. The aim is to identify if the crack propagation instabilities are related to microstructural heterogeneities, i.e. weak zones, or if they are linked to a competition between testing machine stiffness and tearing modulus.An in-depth mechanical instability analysis was performed comparing the testing machine stiffness to the tearing moduli of the irradiated materials. For this purpose an experimental program was carried out in the hot cells facility at CEA Saclay to conduct mechanical and microstructural studies, after irradiation in a Material Testing Reactor (mean conventional thermal fluence: 6.8 × 1021 nth.conv.cm−2).The material hardening was quantified in this study using tensile tests after annealing heat treatments. The effect of this hardening on crack propagation instabilities was measured by fracture toughness tests on irradiated and precracked CT12.5 samples. Damage mechanisms were studied by scanning electron microscope fractography and classical ductile fracture micromechanisms were found for all conditions without notable differences between non irradiated and irradiated materials. The evolution of toughness with irradiation and subsequent heat treatments was therefore attributed to the matrix hardening/softening and its effect on nucleation of voids on brittle second phase particles.To investigate experimentally if the crack propagation instabilities were related to the tearing moduli, annealing heat treatments of the irradiated material were applied that softened the material. It was shown that with softening, the number of crack propagation instabilities decreased and eventually the pop-ins disappeared. The instability criteria using J-Δa R-curves of the toughness tests were applied and it was concluded that the increase in pop-ins with irradiation resulted from the decrease of the material tearing modulus and its interaction with the machine stiffness.The toughness test results of the irradiated material of the present study were compared with those of the literature and show very good agreement.

Scaled-up production of 131I radioisotope using dry distilation method for radiopharmaceutical application

The development of nuclear medicine is increasing every year, including the need for diagnostic or therapeutic radioisotope sources. Iodine-131 (131I) radioisotope is used for diagnostic and therapeutic purposes in patients with thyroid cancer or neuroblastoma. Research and Technology Center for Radioisotope and Radiopharmaceutical - National Research and Innovation Agency (RTCRR-NRIA) has been able to produce 131I radioisotope with radioactivity up to 18.5 – 29.6 Gbq/month. This study aims to scale-up the production capacity of 131I radioisotope by using dry distillation method and to determine the radionuclide and radiochemical purity of the 131I radioisotope. The scaled-up production of 131I radioisotope was carried out in a hot cell facility of Radioisotope and Radiopharmaceutical Laboratory (RRL). 50 grams of TeO2 was used as target material. The TeO2 was irradiated at the Central Irradiation Position (CIP) facility at G.A Siwabessy multipurpose reactor. The post-irradiation target was heated up to 800° C for ± 5 hours under a vacuum condition in the distillation system of -2 incHg to -3 incHg. In this study, the 131I radioisotope was produced with radioactivity of 192 – 344 Gbq/month, radionuclide purity of 99%, and radiochemical purity of 95%. The scaled-up production 131I radioisotope has been successfully carried out with an increase of more than 12 times the initial capacity. In the future, the production of 131I radioisotope will be used for domestic needs in the oral dosage form.

Structural integrity investigations of spent nuclear fuel with finite element modeling

Abstract. Radioactive waste in Switzerland will be disposed of in a deep geological repository (DGR). Responsible for the planning and preparation of realization of this task is National Cooperative for the Disposal of Radioactive Waste (Nagra). Spent fuel assemblies (SFA) constitute the main high-level waste (HLW) stream that will be disposed in the DGR. Prior to final disposal they will be transferred or transported to an encapsulation plant, where they will be loaded into final disposal canisters. To ensure that the structural integrity of SFAs is not compromised during handling and transportation, it is desirable to characterize the expected mechanical parameters of SFAs after long-term interim storage. Experimental research activities performed at the JRC Karlsruhe include safety aspects of radioactive waste management, encompassing also spent fuel storage and spent fuel/HLW disposal activities. Nagra and JRC have established a collaboration to jointly study relevant properties and behaviours of spent fuel rods, with the support of the Gösgen nuclear power plant and of Framatome, and in collaboration with other partners in Europe and internationally. As part of this collaboration, 3-point bending and impact tests were performed at the hot-cell facilities of JRC Karlsruhe, to determine the mechanical response of spent fuel rodlets under quasi-static and dynamic loads. The structural integrity of fuel rods was also evaluated under different handling scenarios using finite element (FE) analysis. Starting with the construction of a static 3D FE model of a Pressurized Water Reactor (PWR) nuclear fuel rodlet in ANSYS Mechanical, Nagra has developed a series of FE models over the years. Mechanical properties of the original rodlet model were derived through an extensive validation process, using experimental data from the 3-point bending tests. To evaluate the mechanical response of an SFA in different loading scenarios, this model was expanded using 1D beam modeling approach. The development of the simplified 1D models is shown in this presentation. In particular, the effect of the contact formulation between the spacer grid and the rods is discussed. Finally, preliminary results of the bending response of a 15×15 PWR SFA sub-model are presented.

Inerting Strategy for a Demonstration-Scale Hot Cell Facility Based on Experiences from Pilot-Scale Argon Cell Facility Operation and CFD Analysis

Pyroprocessing is being developed at Korea Atomic Energy Research Institute (KAERI), and in recent years, all process equipment required for integrated processes have been examined in the PyRoprocess-integrated Inactive DEmonstration (PRIDE) facility. Based on the successful operation of a pilot-scale facility, a conceptual design for this scale-up facility was actualized. Implementing a “demonstration-scale” hot cell facility is challenging as it is intended to supersede PRIDE and satisfy the increased requirements of larger-scale facilities. This study focused on an inerting strategy for a larger-scale (demonstration-scale) hot cell facility to achieve conditions equivalent to those in a pilot-scale gas-tight argon cell facility. The study applies the inerting strategy to a demonstration-scale hot cell facility beyond that of the currently existing pilot-scale hot cell facility and performs computational fluid dynamics (CFD) simulation with various flow rates to determine an appropriate approach for inerting the target facility. To this end, practical constraints on the simulation are introduced based on experiences from the existing pilot-scale facility. The results show that the purging flow rate should be accurately predicted, and a variable flow rate should be applied to achieve hot cell inerting effectively. The required purging time and amount of inerting source are essential factors in the larger-scale hot cell facility. The study results can be helpful in the design of large hot cell facilities operated under inert conditions.

Virtual reality platform-based conceptual design and simulation of a hot cell facility

Virtual reality platform-based conceptual design and simulation of a hot cell facility