Post-irradiation Examination Research Articles

Postirradiation examination results of several metallic fuel alloys and forms from low burnup AFC irradiations

Several different U based metallic nuclear fuel alloys and forms were irradiated as part of the Advanced Fuels Campaign (AFC) series of irradiation experiments. Results from postirradiation examinations of the AFC-3A and AFC-3B are presented here. In this irradiation, U alloyed with Mo, Zr, and Zr with Pd were investigated. Different geometric forms (solid, annular), smear densities and bonding media (sodium, helium) were also irradiated and evaluated. All fuel forms were irradiated in ferritic-martensitic HT-9 cladding. Examination of the irradiated fuel included neutron radiography, evaluation of the distribution of gamma-ray emitting radionuclides, dimensional inspection, fission gas release analysis, optical microscopy, and chemical analysis. No in-pile failure was observed with this fuel. Neutron radiography, gamma spectrometry, and dimensional inspections indicated the fuel largely performed acceptably. Optical microscopy revealed significant fuel-cladding interaction between the Mo based alloys and the cladding. The Zr based alloys generally performed better although some forms revealed behaviors that present specific fabrication challenges such as tighter fabrication tolerances that will need to be better controlled to match historical U-Zr fuel performance. The addition of Pd as an additive to prevent fuel cladding chemical interaction (FCCI) also appears to present some challenges to fuel performance.

Fabrication and characterization of (Th, Pu)O2 fuel at Canadian Nuclear Laboratories

Thorium and other advanced fuels are of interest to the nuclear industry because of their potential role in ensuring long-term sustainable nuclear energy. Thoria (ThO2)-based fuel cycles are options for both conventional water-cooled reactors (including light water and heavy water reactors) as well as various advanced reactor concepts. Plutonium-bearing thoria-based fuels are an attractive option for consuming stockpiles of Pu, while breeding fissile U-233 and extracting energy from thorium. Canadian Nuclear Laboratories (CNL; formerly Atomic Energy of Canada Limited, AECL) has been carrying out research and developing thoria-based fuels, including (Th, Pu)O2, for more than 55 years. The CNL irradiation test of (Th, Pu)O2 fuel assemblies (BDL-422) in the National Research Universal (NRU) research reactor at the Chalk River Laboratories culminated in the publication of post-irradiation examination results during 2008–2013. More recently, CNL has undertaken efforts to further characterize as-fabricated and irradiated (Th, Pu)O2 fuel by developing and implementing improved ceramographic techniques and conducting fabrication trials to better control and improve pellet microstructure. In addition, autoradiography techniques have been developed and applied to irradiated (Th, Pu)O2 fuel to better characterize fuel microstructural behaviour in combination with scanning electron microscope (SEM) characterization. This work has resulted in improvements in the microstructural characterization and fabrication of (Th, Pu)O2 fuels.

Digital neutron and gamma-ray radiography in high radiation environments with an MCP/Timepix detector

The radiation levels from used nuclear fuel make their post irradiation examination very difficult. Handling material removed from assemblies is almost exclusively performed inside shielded hot cells. Neutron radiography provides a non-destructive option for examining the internal structure, location and integrity of irradiated fuel. This paper addresses the performance and stability of a neutron imaging detector comprised of microchannel plates (MCPs) and a Timepix readout in gamma and neutron radiation environments. The detector is being considered as an alternative to the currently used activation foils and photographic film measurement techniques. The MCP/Timepix detector was used in neutron radiography experiment on a beam line at the Neutron Radiography Reactor (NRAD) at Idaho National Laboratory (INL). The detector was installed ∼17.5 m from the reactor core with a line of sight view to the core, which results in a fast and epithermal neutron spectrum as well as high intensity gamma radiation. The gamma-ray dose rate in the beam line was 812 mSv/hr. Complementary measurements were made using 137Cs and 60Co isotopic sources at gamma dose rates in excess of 6 Sv/hr. The detector functioned normally over several hours in this environment. These conditions are comparable to planned radiography of irradiated nuclear fuel assemblies. Experimental results using ASTM sensitivity and beam purity indicators suggest that a spatial resolution of <100 μm may be possible in neutron radiographic measurements using this detector for studies of irradiated nuclear fuel. Potential radiographic and tomographic applications are discussed. The results of present experiments are not limited to the studies of irradiated fuel assemblies, but rather are relevant to any experiment where imaging needs to be performed in a high radiation environment.

Neutron propagation experiments with a lead test section inserted in the core of the LR-0 reactor

Validation of neutronic parameters for lead-cooled systems, by integral and local experiments, became of primary interest with advances in ALFRED lead cooled reactor demonstrator design. Notably, aiming at advancing the core design from the conceptual to the basic stage, the assessment of uncertainties coming from numerical simulation, with respect to real experiments, was of paramount importance to support the claims of outstanding safety demonstration meant for ALFRED. Among the required evidences, the assessment of the spatial distribution of the neutron flux and power in fuel pins required conceiving ad-hoc experiments, and disposing of state-of-the-art post-irradiation examination capabilities. For all these reasons, these experiments were one of the main focuses of the collaboration between the Research Centre Rez and ENEA. This paper presents the conception phase of the test and discusses some of the main results of the first phase of the experimental campaign, dealing with neutrons propagation, collected during its execution at the LR-0 research reactor (hence before the post-irradiation examination of the experimental pins placed in the lead test rig). The experimental work involved at first neutron spectrum measurements in the energy range from 0.1 to 10 MeV. Additionally, measurement of basic neutronic parameters of lead were performed, such as its reactivity worth, its effect on the neutron spectrum and its slowing-down properties. The comparison of calculations and experimental results shows good agreement. In case of calculation in benchmark model with different nuclear data libraries, the criticality is systematically over-predicted by approximately 150 pcm, which is, however, in the 1σ of the uncertainty interval. Neutron spectrum measurement shows only slight variations being around 10% most of the time.

Development and qualification of functional materials for the European HCPB TBM

The paper reviews the current status of development and qualification of the HCPB TBM’s functional materials, namely, Li-ceramic breeder and Be multiplier materials. The main functional and performance requirements for both functional materials are overviewed. The main results and outcomes of the post-irradiation examinations (PIE) of Li-ceramic breeder and Be/beryllides materials irradiated in HIDOBE and HICU irradiation campaigns are presented aiming at determination of the materials’ performance, properties and characteristics under neutron irradiation, like e.g. morphology and microstructure, tritium retention/release characteristics, mechanical properties. Fabrication of the advanced ceramic breeder material, containing lithium orthosilicate with addition of lithium metatitanate phase in order to improve its mechanical properties, is described. Results of characterization of this advanced ceramic breeder in terms of mechanical properties, porosity, morphology and effective thermal conductivity of a pebble bed are overviewed. In addition, results of characterization of the reference 1 mm Be pebbles produced by Rotation Electrode Process are presented in terms of their microstructure, tritium release characteristics and interaction with air and steam. A necessity to realize in the future a new irradiation experiment for the functional materials, LIBERTI experiment, is discussed in the conclusions.

An atomistic study of grain boundaries and surfaces in γU-Mo

A monolithic fuel design based on a U-Mo alloy has been selected as the fuel type for conversion of the United States High-Performance Research Reactors (HPRRs). A 2015 post-irradiation examination (PIE) report showed accelerated swelling in U-10Mo fuels at fission densities much lower than previously observed. This PIE report showed a large amount of compositional banding, or regions of low Mo content adjacent to regions of high Mo content, with low Mo content typically along grain boundaries. Lower Mo content can lead to phase decomposition from the gamma U-Mo body-centered cubic phase to the alpha U phase as well as an earlier onset of recrystallization. Thus, the phenomenon of Mo depletion at grain boundaries is an important factor in the accelerated swelling behavior of U-Mo fuel. However, the physical origin of Mo depletion at grain boundaries is still unclear. In this work, molecular dynamics simulations have been performed to calculate the grain boundary and surface energies of body-centered cubic (bcc) U, bcc Mo and alloys of U-Mo from 600 K to 1200 K. It is observed that the lower grain boundary energy of bcc U, compared to bcc Mo, provides the driving force for Mo depletion at grain boundaries. This driving force diminishes with increasing temperature, but is not eliminated. This information can be utilized as inputs to higher length scale modeling methodologies and provide specification guidance to fabricators.

Performance of U3Si–Al dispersion fuel at HANARO full-power condition

Abstract The irradiation performance of U 3 Si dispersion fuel in an Al matrix, U 3 Si-Al, under the Hi-Flux Advanced Neutron Application Reactor (HANARO) design full-power condition of 30 MW was tested for full-power qualification of the fuel. A test assembly was fabricated containing 18 fuel rods made with atomized U 3 Si powder manufactured at the Korea Atomic Energy Research Institute. The test assembly was irradiated for 188 full-power operation days in the HANARO subject to the normal fuel-loading scheme and achieved about 60 at% U-235 average burnup and 75 at% U-235 peak burnup. The maximum linear power of the test assembly was 98 kW/m. Nondestructive and destructive postirradiation examinations were conducted. The measured postirradiation examination data were compared with data from previous irradiations and the design criteria required for HANARO fuel. Consequently, it was concluded that in-pile performance was acceptable and fuel integrity was maintained, and the behavior satisfied the fuel design requirements.

Data collection assessment for the human performance analysis in nuclear installations

Abstract The purpose of this paper is to provide a good knowledge and understanding on the main factors influencing human performance in complex systems operation. Identification and establishment of the necessary data for human factor performance analysis in a socio-technical context based on the analysis of HRA methods are performed. An evaluation of the techniques used for collecting and characterization of necessary data for HRA (observation, questionnaire, interview, task analysis, change analysis, simulator, expert judgement) is achieved also. The necessary data to perform HRA for the TRIGA reactor operation, for activities of the post-irradiation examination laboratory and activities of the radioactive waste plant treatment from RATEN ICN Pitesti were obtained. Recommendations on how to obtain data and knowledge related to development of methodical framework of HRA data collection, characterization and evaluation are also presented in this paper.

Small specimen test techniques for evaluation of tensile flow properties – Evolution and developmental activities at IGCAR, Kalpakkam

The small specimen test techniques have been a subject of active research for the past 30 years. The idea of using small or miniature specimens for mechanical testing had actually originated in nuclear industry caused by space limitations in nuclear reactors for irradiation material testing with the added advantage of reduced radiation dose from smaller volumes during post-irradiation examination. Considering the potential applications of this technology for material characterization, the programme was initiated at IGCAR with the ab initio development of non conventional test techniques based on the use of small disc specimens of 3–8 mm diameters and 0.3–1.0 mm thick. The techniques that have been extensively studied are shear punch, small punch and ball-indentation tests. This paper will present the evolution of the research and developmental activities on punch and indentation based test techniques at the authors’ laboratory along with their salient results and highlight the outcome of few applications of small specimen tests employed for evaluating tensile flow properties.

The first irradiation testing and PIE or TRISO-coated particle fuel in Korea

The first irradiation testing of TRISO-coated particle fuel to demonstrate and qualify its use in a VHTR was finished at the beginning of 2014 in HANARO, Korea. A PIE (post-irradiation examination) of the irradiated TRISO-coated particle fuel was then performed. The irradiation device contained two kinds of test rods: one containing nine fuel compacts, and the other five compacts and eight graphite specimens between the fuel compacts. The irradiation testing was carried out under the condition of filled with an inert gas atmosphere and being sealed (no sweep gas). After irradiation testing, the maximum power of the fuel compact was estimated to be around 56 W at 25 EFPD, and the maximum power of the particle was calculated to be 215.4 mW. And, the maximum discharged burn-up was evaluated to be about 37,344 MWD/MTU (3.99% FIMA). The estimated maximum fast fluence of the graphite specimen was 2.99 × 1020 n/cm2 (E > 0.18 MeV). In this paper, the characteristics of TRISO-coated particle fuels, the estimated thermal power, fuel temperature, burnup, fluence, and some PIE results were described.

Shutdown-induced tensile stress in monolithic miniplates as a possible cause of plate pillowing at very high burnup

Post-irradiation examination of RERTR-12 miniplates showed that in-reactor pillowing occurred in at least 4 plates, rendering performance of these plates unacceptable. To address in-reactor failures, efforts are underway to define the mechanisms responsible for in-reactor pillowing, and to suggest improvements to the fuel plate design and operational conditions. To achieve these objectives, the mechanical response of monolithic fuel to fission and thermally induced stresses was modeled using a commercial finite element analysis code. Calculations of stresses and deformations in monolithic miniplates during irradiation and after the shutdown revealed that the tensile stress generated in the fuel increased from 2 MPa to 100 MPa at shutdown. The increase in tensile stress at shutdown possibly explains in-reactor pillowing of several RERTR-12 miniplates irradiated to the peak local burnup of up to 1.11 × 1022 fissions/cm3. This paper presents the modeling approach and calculation results, and compares results with post-irradiation examinations and mechanical testing of irradiated fuel. The implications for the safe use of the monolithic fuel in research reactors are discussed, including the influence of fuel burnup and power on the magnitude of the shutdown-induced tensile stress.

Identification of the contributors (Ag-110 m) for higher radiation field on primary heat transport system of Tarapur Atomic Power Station-3 and its impact on collective dose

Tarapur Atomic Power Station-3 is 540 MWe pressurized heavy water reactor in India, it achieved first criticality on May 5, 2006 and then operated at full power. The control of dose rates and collective dose of radiation worker is most important for the best performance of reactor. This article discusses the sudden rise in radiation levels on primary heat transport (PHT) pipelines and equipment. The radionuclide contributed to high radiation levels was Ag-110 m. Finding obtained during the route cause analysis used for the station operation and removal of Ag-110 m from PHT system using special type of resin (macroporous). The purification of the system using special type of resin has been enhanced. Ag-110 m activity in the PHT fluid has been controlled. PHT pipelines and equipment shielded for exposure control. The postirradiation examination of steam generator (SG) manhole gasket showed the presence of sulfur. The PHT fluid sample analysis revealed the presence of oil content in PHT fluid which was ingressed from buffer tank and FT D2O tank in PHT system. Oil containing sulfur was responsible for erosion of silver from SG manhole gasket in the form of silver sulfide and its ultimate transport to reactor core and neutron activation of silver.

HTR-PM fuel pebble irradiation qualification in the high flux reactor in Petten

NRG has performed a High Temperature Reactor fuel qualification test in the framework of the Chinese HTR development for INET – Tsinghua University, as a part of the qualification process of Chinese fuel for use in the HTR-PM (High Temperature Reactor-Pebble bed Module). Fuel qualification is an essential part of the licensing process of the HTR-PM, currently under construction in the Shandong province in the People’s Republic of China.The design of the INET irradiation setup for irradiation in the High Flux Reactor is performed by NRG. In the design of the irradiation rig, the pebbles are encased in graphite half shells to form a solid cylindrical stack. Within the graphite, thermocouples are located for temperature monitoring, and neutron fluence monitor sets for neutron fluence verification. Next to the temperature measurements, the online monitoring consisted of fission product release measurements. These measurements were performed with the Sweep Loop Facility, designed and built by JRC-IET.Post Irradiation Examinations are performed and consisted of gamma scanning, dimensional and weight measurements and visual inspection. In April 2016, the pebbles were transported from Petten to Karlsruhe, Germany where the second required step of fuel qualification will be performed, namely heating tests in the KÜFA facility.This paper describes the global lay-out and functionality of the irradiation facility. The online data generated and irradiation parameters are reported, with focus on the measured fission product release behavior of the fuel. The paper also presents non-destructive post irradiation examinations.

The AGR-3/4 fission product transport irradiation experiment

AGR-3/4 was the combined third and fourth planned irradiations for the U.S. Department of Energy (DOE) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The primary purpose of the AGR program is to support the development and qualification of tristructural isotropic (TRISO)-coated particle fuel for use in High Temperature Gas-cooled Reactors. AGR-3/4 was designed as a fission product transport irradiation experiment whose specific objectives were to: (1) irradiate fuel containing UCO (uranium oxycarbide) designed-to-fail (DTF) fuel particles that provide a fixed source of fission products for subsequent transport through compact matrix and structural graphite materials; (2) assess the effects of sweep gas impurities on fuel performance and fission product transport; (3) provide irradiated fuel and material samples for post-irradiation examination (PIE) and post-irradiation heating; and (4) support the refinement of fuel performance and fission product transport models. The AGR-3/4 test train was irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for 369.1 effective full power days from December 2011 to April 2014. The experiment was successful in achieving its specification goals in terms of burnup and fast fluence levels reached at the end of irradiation and fuel temperature levels maintained throughout irradiation: peak compact burnup reached 15.27% fissions per initial heavy-metal atom and peak compact fast fluence reached 5.32 × 1025 n/m2 (E > 0.18 MeV), while the time-average volume-average temperatures of the compacts ranged from 854 to 1345 °C. Fission product release–to–birth ratios reached values in the 10−4–10−3 range early during irradiation as the DTF particles started to fail. Subsequent post-irradiation examination will provide information on fission product distributions in matrix and core graphite materials, enabling refinement of fission product transport models.

Stability of the structure and properties of an ultrafine-grained Cr-Ni steel irradiated with neutrons in nuclear reactor core conditions

An ultrafine-grained (UFG) austenite Cr-Ni stainless steel was produced by equal-channel angular pressing. The UFG samples were irradiated in a research nuclear reactor up to damaging doses of 12 and 15dpa at 350°C and 450°C, respectively. The post-irradiation examination featured in this work covers microstructural and mechanical property changes at test temperatures of 20–650°C. Microstructural studies showed that the concentration of Frank loops was considerably lower in the UFG steel compared to in the coarse-grained material. It was also found that irradiation resulted in precipitation of α-phase in triple junctions of the UFG steel. The post-irradiation examination of mechanical properties testified that the UFG steel exhibited lower irradiation hardening and higher strength compared to the coarse-grained material in the entire testing temperatures range, and improved ductility when tested at 500–650°C. In summary, we testified that microstructure refinement of an austenite stainless steel down to UFG scale led to enhancement of its resistance to neutron irradiation at elevated temperatures.

Evaluating the irradiation effects on the elastic properties of miniature monolithic SiC tubular specimens

The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. The optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.

Initial examination of fuel compacts and TRISO particles from the US AGR-2 irradiation test

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second irradiation test. These compacts were selected for examination because there were indications that they may have contained particles that released cesium through a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzer (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using X-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and X-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers.

Conceptual design of test devices for the JHR tailored to the needs of the nuclear fusion community

Since 2015, within the scope of the Jules Horowitz Reactor (JHR) project a collaborative effort between CCFE in the UK and CEA in France has been undertaken to investigate concepts of test devices for the JHR suitable for the needs of fusion researchers. It is hoped that in having optimized experimental devices designed will facilitate use of the JHR by the fusion community.The project chose to focus on more instrumented test devices as it was felt that post-irradiation examination (PIE) type experiments at conventional temperatures could be undertaken with limited modifications to existing test devices. After internal exploration of the options, three potential test devices were identified; i) testing of ceramic functional materials; ii) testing at cryogenic temperatures; and iii) testing of thermo-mechanical fatigue. Each of these devices has been conceptually designed, demonstrating feasibility.This paper describes each of the conceptual designs at their current level of maturity. Future collaborative work planned between the two parties will aim to develop these devices further.

Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700°C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.

Post-Irradiation Non-Destructive Analyses of the AFIP-7 Experiment

This article reports the results and interpretation of post-irradiation non-destructive examinations performed on four curved full-size fuel plates that comprise the AFIP-7 experiment. These fuel plates, having a U-10 wt.%Mo monolithic design, were irradiated under moderate operating conditions in the Advanced Test Reactor to assess fuel performance for geometries that are prototypic of research reactor fuel assemblies. Non-destructive examinations include visual examination, neutron radiography, profilometry, and precision gamma scanning. This article evaluates the qualitative and quantitative data taken for each plate, compares corresponding data sets, and presents the results of swelling analyses. These characterization results demonstrate that the fuel meets established irradiation performance requirements for mechanical integrity, geometric stability, and stable and predictable behavior.