Fission Product Release Research Articles

Study of flammable gases generation and radionuclide release during and after drying of damaged VVER SNF

The paper presents an experimental evaluation of drying the damaged VVER SNF after the wet storage. The aim of the experiments was to justify the fire safety of air-tight canisters with the dried SNF during transportation, storage, and reprocessing. Keeping the SNF in boric acid during the experiments determined the concentration of uranium, plutonium, and specific activity of fission products in the solution inside the model canisters with the SNF. The canisters with the SNF underwent the thermal vacuum drying to determine the process parameters (temperature, residual pressure, drying rate), a hydrogen release, 85Kr activity during the drying, and an airborne radionuclide release onto the filters of the SNF thermal vacuum drying system. The results of weighing the baskets of the canisters before soaking in the boric acid and after drying at 160 °С to the residual pressure below 6 mbar showed that the difference in weight is almost the same as the calculated difference in weight of H3BO3 loaded into the canister and the weight after the complete decomposition into B2O3. Thus, we have discovered that the boric acid undergoes complete decomposition into boric anhydride, and the SNF corrosion products generate hydrogen that accumulates in the air-tight canister. We also considered the hydrogen that accumulated in the air-tight model canister with the dried SNF as a result of the radiolysis of hydrated corrosion products of the spent oxide nuclear fuel, as well as a release of gaseous fission products (GFP) inside the canister. We calculated the quantity of the hydrated corrosion products of uranium oxide after the wet storage and determined their chemical composition after the thermal-vacuum drying. The equation with an exponential decay function describes the hydrogen accumulation. The calculated maximum volume of hydrogen and uranium in the SNF corrosion products suggests that the SNF corrosion product is schoepite with the number of gyrated water molecules corresponding to the formulas from UO3·0.95H2O to UO3·1.5H2O. The received data allows for the fire safety justification of drying the wet damaged SNF and handling the dried SNF during transportation and storage.

Uncertainty and sensitivity analysis of the fission product behaviour in the Phébus FPT1 test with the system code AC2

The Phèbus FP (Fission Product) research program, conducted by IRSN and CEA, aimed to improve the understanding of the phenomena governing core melt down, fission product release and transport, as well as the fission product behaviour in the containment. The test facility represents a 900 MWe PWR at a 1/5000 scale. The second test, FPT1, was carried out in 1996 and utilised an Ag-In-Cd alloy control rod. There is a growing interest in Best Estimate Plus Uncertainty analysis both in the research, as well as in the licensing. It has a long-standing history in case of thermal–hydraulic system codes focusing on design basis accidents, but its application is limited on the severe accident domain. In recent times however an increasing attention has been paid to evaluating severe accident codes’ uncertainties. This study presents an uncertainty and sensitivity analysis (UaSA) of the FPT1 test, focusing on fission product behavior modeling within the severe accident code system AC2. The detailed application of the coupled codes ATHLET-CD and COCOSYS in this context is novelty. Detailed description of the methodological and practical approach is provided. The uncertain input parameters chosen and quantified are directly related to the phenomena describing the fission product transport and retention phenomena both in the primary circuit, as well as in the containment, while thermal–hydraulic conditions within the primary circuit were kept as much as possible constant. Finally, the study has been complemented by a sensitivity analysis deriving the Spearman's rank correlation coefficient for each uncertain input parameter.

Coupled computational fluid dynamics and computational thermodynamics simulations for fission product retention and release: A molten salt fast reactor application

This study presents a computational capability for fission product retention and release in two-phase, multi-species systems representing Molten Salt Reactors (MSR) with coupled thermal-hydraulics and fuel coolant chemical behaviours. This is demonstrated through four simulated cases centred on the proposed Molten Salt Fast Reactor (MSFR). This is achieved by two-way coupling the Computational Fluid Dynamics (CFD) code OpenFOAM and the Computational Thermodynamics (CT) code Thermochimica, using the Joint Research Centre Molten Salt Database (JRCMSD). Local chemical equilibrium is assumed, implying that chemical kinetics are predominantly governed by mass transport. Four simulations address normal operating conditions, exploring: (i) dilution of fission products injected within the molten salt coolant, (ii) molten salt coolant evaporation rate, (iii) release of radioactive gaseous species, (iv) shifts in the UF4/UF3 ratio, and (v) comparison of vapour pressures of gaseous species. The influence of temperature-dependent viscosity on retaining fission products, compared to consistent values, is also discussed. The feasibility of integrating CFD with Thermochimica showed promising results, broadening insights into multiphysics systems and setting the stage for its application in more intricate scenarios.

Simulation of late phase phenomena with the system code AC2

Simulation codes like the AC2 system code package developed by German Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH are used in international reactor safety research to investigate hypothetical severe incidents and accidents in nuclear power plants. Valuable impetuses for the development of prevention and mitigation measures for such events can be derived from the analyses. This article presents selected work by PSS on the development of simulation models for the late phase of nuclear accidents. Topical issues as the formation and thermodynamic interactions of debris beds in the lower plenum of the reactor pressure vessel (RPV), the release of fission products as well as the interaction of the molten core material with concrete in the reactor cavity in case of RPV failure are discussed. The analyses carried out provide important findings regarding the validation of the developed models. In addition, future development needs are derived for the consideration of further relevant phenomena.

Impact of temperature variations on BISON predictions of Ag release in AGR-1 and AGR-2 experiments

Understanding and quantifying the release of fission products like silver (Ag) from TRistructural ISOtropic (TRISO) fuel particles is important to assess the safe operation of advanced high temperature reactors. Although the silicon carbide (SiC) layer of TRISO particles is effective as the main fission product barrier, Ag can be released from intact TRISO particles. A mechanistic model for the effective Ag diffusivity, Deff, was previously developed as a function of temperature and microstructure variables informed by atomistic modeling of Ag diffusivity. In this study, we use this model to explore how experimental temperature uncertainties impact the overall predicted Ag release. This analysis indicates that calculated temperature uncertainties have a significant impact on the overall Ag release predictions. Furthermore, we show that the time average volume average (TAVA) temperature is not an appropriate proxy for the detailed temperature histories to predict fission product release due to the Arrhenius dependence of Ag diffusivity with respect to temperature. The detailed temperature histories, therefore, provide the most accurate results and are of most importance for modeling efforts. Overall, this work shows the importance of considering the experimental uncertainty of the temperature on computational predictions of fission product transport and release and the need for more accurate temperature histories from future experiments.

Flow visualization experiments of argon injection in a molten salt natural circulation loop

Off-gas systems are implemented in molten salt reactor designs to control the release of gaseous fission products. Two-phase flow in molten salt must be studied to understand how the system will behave in comparison to traditional working fluids like water. Flow visualization experiments and particle image velocimetry measurements were performed for three argon bubble sizes injected into a co-current stream of molten salt in a natural circulation loop facility. Similar bubble sizes were injected in experiments with water to compare the bubble shape, trajectory, and wake flow behavior of the fluids. The bubble region of interest was used to calculate the equivalent diameter and terminal velocity. Results for water showed a wobbling bubble surface and less stable bubble trajectory due to lower surface tension and viscosity compared with molten salt. Particle image velocimetry results demonstrated the increased viscosity of salt dampens turbulent fluctuations for the smaller bubble size. For a cap bubble, turbulent fluctuations were larger and longer lasting than in results for the wake flow of an argon cap bubble in water.

Source term analysis of FeCrAl accident tolerant fuel using MELCOR

It has been established that incremental improvements in beyond design basis accident performance can be achieved through accident tolerant fuel (ATF). However, they have the potential to recover margin with respect to conventional fuel and therefore enhance plant economics through uprate or cycle length increase. To realize this potential, it is necessary to quantify the reduction in source term due to use of ATF, and correspondingly how this is affected by increasing the cycle length and/or burnup. This requires development of a risk informed analysis methodology for ATF under high burnup conditions, which is being developed within LWRS. To this end, a MELCOR simulation of FeCrAl ATF in a recovered Large Break LOCA (LBLOCA) scenario has been developed, using a model based on the Zion Pressurized Water Reactor (PWR). The new user defined material capability, along with the inclusion of detailed neutronics- and depletion-derived parameters such as core power distribution, decay heat behavior, and fission product inventories, allows a more detailed simulation of FeCrAl clad material properties and behavior than has previously been possible using MELCOR. These detailed FeCrAl results were compared with a zircaloy clad model to investigate the differences between the two clad types and quantify the benefits of FeCrAl ATF with an 18-month cycle. Next, the fuel cycle with ATF was extended to 24 months, to determine whether any additional safety margin provided by FeCrAl ATF could be leveraged to implement high burnup FeCrAl-clad fuel while retaining the same operating and safety limits as current zircaloy-clad fuel. For the particular scenario analyzed, the delay to fuel failure from using ATF was of the same order as the LOCA recovery time, and hence significant in reducing fission product release, with the 24-month FeCrAl core performing better than the 18-month zircaloy core. It is noted that for other transients, the reduction in release due to using FeCrAl may be less significant. The material model developed here can be used in such further studies in support of determining the overall source term.

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

Fission product release from spent nuclear UOX fuel dissolution: Comparison between anoxic and reducing conditions and impact of pH

Several clad segments from the same spent nuclear fuel rod were leached in a highly alkaline solution representative of the conditions expected in the current Belgian reference deep disposal design, and in presence of hydrogen. The results are compared with experiments that were conducted with clad and declad samples in a bicarbonate solution without presence of hydrogen. For the monitored radionuclides (238U, 90Sr, 137Cs, 129I, 99Tc), a fast initial release was observed. Afterwards, the release depended on the experimental conditions (medium, redox conditions) for uranium, strontium and technetium, but not for cesium and iodine. The cumulated released fractions were quantified at the end of the experiment.Graphical

Validation of the severe accident module of the EVKLID/V2 integral code on the base of experiments with fission products release and dissociation of nitride fuel

Validation of the severe accident module of the EVKLID/V2 integral code on the base of experiments with fission products release and dissociation of nitride fuel

Modelling of defective CANDU fuel phenomena

A mechanistic model is presented for the modelling of defective CANDU fuel behaviour. This physically-based treatment includes a multi-physics coupling of fission product release, fuel oxidation and sheath hydriding phenomena with the occurrence of fuel failure. The stoichiometry deviation profile is used to evaluate effects on the fission-product diffusivity and heat conduction in the fuel. A time-dependent fission product release model accounts for the transport of fission products of the short-lived radioiodine and noble gas isotopes in the fuel matrix with their release into the fuel-to-sheath gap and coolant. Transport of steam and hydrogen gases along the gap is further considered as a boundary condition of the fuel oxidation model. This treatment provides an estimate of the partial pressure ratio of hydrogen-to-steam in the gap for evaluation of the critical ratio necessary for the promotion of sheath hydriding. Secondary hydriding and fuel element deterioration is subsequently evaluated in accordance with a Locke-type correlation.

FISSION PRODUCT RELEASE FROM HIGH AND LOW-ENRICHED URANIUM FUELS OF THE IVG.1M RESEARCH REACTOR

The research involved conducting resource tests on two experimental water-cooled technological channels (WCTCs) utilizing low-enriched uranium (LEU) fuel within the IVG1.M research reactor. This testing was a crucial step in the reactor's conversion from highly enriched uranium (HEU) fuel to LEU. The research focused on two key parameters, namely the specific activity and the relative release of fission products (FPs) into the coolant, to evaluate the tightnessof the fuel element cladding.A gamma-spectrometric sampling method was proposed to determine the relative releaseof FPs, which involved assessing the specific activity of the coolant, calculating the release rate (Release), the born rate (Born), and the R/B ratio of FPs. Comparative gamma-spectrometric measurements were conducted to analyze the content of FPs and activation products (AP) in the coolant of WCTCs utilizing both LEU and HEU during the tests.From the comprehensive list of detected radionuclides in the IVG.1M reactor coolant, well-identified reference radionuclides recommended for monitoringfuel element cladding tightness were carefully selected. The results of the study provided insights into the specific activity and relative release of FPs, demonstrating that quantitative values for the relative release of FPs from WCTCs using LEU and HEU fuel were comparable.

Performance evaluation of an improved pool scrubbing system for thermally-induced steam generator tube rupture accident in OPR1000

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

Towards modelling defective fuel rods in TRANSURANUS: Benchmark and assessment of gaseous and volatile radioactive fission product release

This work presents the results of a collaborative benchmark activity between different organizations towards the use of the TRANSURANUS code to estimate the release of gaseous and volatile radioactive fission products from defective fuel rods, into the primary coolant of pressurized water reactors. First, the radioactive release from the fuel to the gap is evaluated according to three approaches: the coupling between TRANSURANUS and SCIANTIX, the development of TRANSURANUS devoted subroutines, and the use of the ANS 5.4–2010 methodology. Fuel-to-gap release calculations are benchmarked and assessed against measured data from the CONTACT1 irradiation experiment. Then, TRANSURANUS has been used to estimate the radioactive release into the primary coolant by applying a first-order phenomenological rate theory and tested against measured data of fission product coolant concentrations from irradiation experiments of the CRUSIFON program.

Iodine source term assessment under DBA SGTR accident scenario

In the framework of the European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA), application of Best Estimate advanced codes for evaluating the radiological release under Design Basis Accident and Design Extended Conditions was considered. The aim is to assess the current modelling capabilities and to propose new calculation methodologies in order to produce more realistic evaluations of radioactive releases resulting from such accidents.In this paper, the studies are conducted for a reference Design Basis Accident Steam Generator Tube Rupture scenario in a 3-loop PWR-1000 reactor in terms of key thermal–hydraulic and radiological variables that condition and characterize the fission product release to the environment. The assessment includes a comparative study of the results performed by three organisations using three different computational tools and approaches. The outcomes highlighted issues considered for the design basis evaluation, modelling differences as well as some challenges and limitations in carrying out such analysis.

Development and validation of the MAVR-TA code for analyzing the release and transport of fission products during a severe accident at a VVER NPP. Part 3 – Modelling of the release of fission products from the melt

The MAVR-TA code is designed for modelling of fission product release from fuel and melt, as well as fission products transport in the primary circuit and in the containment during a severe accident at a NPP with VVER. The third part of the work provides information on the models of the MAVR-TA code related to modelling of fission products release from the melt due to evaporation from the melt surface and due to molten core concrete interaction. The paper describes the models and approximations implemented in the code, as well as experimental tests used to validate the models of fission product release in the MAVR-TA code. The results obtained during validation are described and analyzed.

Effect of hydrogen gas and leaching solution on the fast release of fission products from two PWR fuels

To study the dissolution of UOX spent nuclear fuel in a deep geological environment and the fast release of a selection of relevant radionuclides for long-term safety of this high level waste, leaching experiments were performed with spent nuclear fuel samples originating from the pressurized water reactors (PWRs) Tihange 1 and Gösgen with a similar burnup (50 – 55 MWd.kgHM−1) but different irradiation histories. Six experiments were conducted to investigate the effect of two critical parameters: (1) the highly alkaline environment caused by the presence of cementitious materials in the “Supercontainer design”, which is currently the reference design for the long-term management of the high-level nuclear waste forms in Belgium, and (2) the reducing conditions imposed by the presence of hydrogen from the corrosion of iron-based materials present in the engineered barriers. The experiments were performed using autoclaves under pressure from 1 to 40 bar with a pure Ar atmosphere or a mixture of H2/Ar. Divided into two consecutive phases, the total experimental duration was about 1400 days. The Phase I provided mainly information about the fast release of the fission products while the perspective of the Phase II was to study the long-term evolution of the spent fuel matrix. During the leaching experiment, concentrations of a selection of radionuclides (238U, 129I, 137Cs, 90Sr and 99Tc) were monitored in solution and the amounts of Kr and Xe were measured in the gas phase. Based on results of the experiments conducted for up to 40 months (i.e. during Phase I of the experimental program), we observe that there is a continuous release of 137Cs, 90Sr and of the fission gases and a clear impact of the irradiation history on the release of certain fission products.

Iodine and fission gas release rates from HTR fuel – experiments, correlations, and mathematical models

Measurement and prediction of fission product release is important in design, licensing and operation of HTRs. Latest development status is near zero particle manufacturing defects and in-pile failures combined with very low matrix contamination levels. In-pile Release over Birth rate measurements will be described and evaluated. Quantitative predictive analyses of short-lived gas release will be provided for MTR irradiation testing with a focus on the most recent experiment with Chinese fuel from HTR-PM production. For irradiation tests with no single particle failure, models describing particle kernel release are not applicable; instead modeling of matrix contamination release is essential. Good agreement between postcalculations and in-line R/B measurements has been achieved.

A new design of composite fuel with Zr-coated TRISO particles dispersed in metal matrix

With the development of advanced reactors, it is critical to improve the safety of nuclear reactors. Safety primarily depends on the ability to block fission products and the ability to discharge decay heat of fuel elements. To enhance the ability to retain the release of fission products and achieve high thermal conductivity, a novel fuel pellet was developed, consisting of zirconium (Zr)-coated TRISO particles embedded in a zirconium metal matrix. The zirconium coating was deposited via fluidized-bed chemical vapor deposition (FB-CVD), and the deposition mechanism of Zr coating via FB-CVD was proposed. The consolidation process of Zr-coated TRISO particles in the Zr matrix was achieved via spark plasma sintering (SPS). The microstructure and mechanical properties of Zr coating and the pellet were investigated via SEM and nanoindenter. The results proved the effectiveness of zirconium coating in preventing TRISO fuel particles from contact and cracking during sintering.

Methods for the radioactive release estimation under DEC-A conditions

After the Fukushima Daiichi Nuclear Power Plant accident, the importance was raised to strengthen the assessment of the safety level of Nuclear Power Plants (NPP) by considering situations more severe than those integrated during the design of the plants. Design Extension Conditions (DEC) term was introduced by IAEA and WENRA. One of the most important subjects in the analysis of these conditions is the evaluation of radiological consequences.The Reduction of Radiological Accident Consequences (R2CA) collaborative project started in 2019 in the frame of the Horizon-2020 Program of the European Commission. The project addresses a broad scope of LWR designs (Gen II, III and III +) through the analyses of bounding scenarios of Loss Of Coolant Accidents (LOCA) and Steam Generator Tube Rupture (SGTR) transients and explores DBA and DEC-A conditions. The Lithuanian Energy Institute is taking part in this project as well.The part of work provided in the frame of R2CA project is presented in this article. The generic BWR-4 type reactor with MARK-I containment was analyzed in the case of DEC-A conditions. In this analysis the main thermal hydraulic processes, fission product release and its transport from the broken loop through containment to the environment were investigated. The number of ruptured fuel assemblies is the key parameter determining the fission product release from the core. The methodology which would help more precisely evaluate the number of failed fuel assemblies is presented in this work. The methodology consists of the application of integral severe accident code ASTEC and fuel performance code TRANSURANUS. Analysis using ASTEC code showed the importance of the reactor core nodalisation scheme. Taking iterative calculations, the relative power to which fuel assemblies remain intact at the selected DEC-A condition scenario was obtained. These results were verified with TRANSURANUS code calculations considering the uncertainties of the most relevant parameters. Based on TRANSURANUS calculation results, the ASTEC core nodalisation was improved. Improved ASTEC model nodalisation allowed to calculate more realistic rates of fission products releases from the fuel to the coolant, from the coolant to the containment and as well from the containment to the environment.