Fuel Decay Heat Research Articles

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

Comparison of calculated and measured spent nuclear fuel decay heat with CASMO5, SNF and standard methods

In this paper, we present a comparison between measured and calculated decay values for spent nuclear fuel assemblies from the Clab, GE-Morris EMAD facilities. Two codes are applied: CASMO5 and SNF, using different nuclear data libraries (JEF-2.2, JEFF-3.1.1, JEFF-3.2, ENDF/B-VII.1 and ENDF/B-VIII.0) and different code versions. Additionally values from four standard and guides for calculation of decay heats (ANS 2014, DIN 2014, ISO 2022 and RG-3.54 2018) were also compared with the measurements. All simulations with different codes, code versions and methods are based on the same geometry and irradiation descriptions for a single assembly, providing more convenient comparisons of results. The impact of nuclear data libraries and computational methods are shown, compared to the spread of C/E (Calculations over Experiment) values between the 256 considered measurements. Finally, a comparison of the most important nuclides contributing to the decay heat is presented between measured assemblies and more recent cases corresponding to higher burnup, initial enrichment and longer cooling times.

Analyses of the bias and uncertainty of SNF decay heat calculations using Polaris and ORIGEN

The bias and uncertainty of calculated decay heat from spent nuclear fuel (SNF) are essential for code validation. Also, predicting these quantities is crucial for deriving decay heat safety margins, influencing the design and safety of facilities at the back end of the nuclear fuel cycle. This paper aims to analyze the calculated spent nuclear fuel decay heat biases, uncertainties, and correlations. The calculations are based on the Polaris and ORIGEN codes of the SCALE code system. Stochastically propagated uncertainties of inputs and nuclear data into calculated decay heats are compared. Uncertainty propagation using the former code is straightforward. In contrast, the counterpart of ORIGEN necessitated the pre-generation of perturbed nuclear cross-section libraries using TRITON, followed by coincident perturbations in the ORIGEN calculations. The decay heat uncertainties and correlations have shown that the observed validation biases are insignificant for both Polaris and ORIGEN. Also, similarities are noted between the calculated decay heat uncertainties and correlations of both codes. The fuel assembly burnup and cooling time significantly influence uncertainties and correlations, equivalently expressed in both Polaris and ORIGEN models. The analyzed decay heat data are highly correlated, particularly the fuel assemblies having either similar burnup or similar cooling time. The correlations were used in predicting the validation bias using machine learning models (ML). The predictive performance was analyzed for machine learning models weighting highly correlated benchmarks. The application of random forest models has resulted in promising variance reductions and predicted biases significantly similar to the validation ones. The machine learning results were verified using the MOCABA algorithm (a general Monte Carlo-Bayes procedure). The bias predictive performance of the Bayesian approach is examined on the same validation data. The study highlights the potential of neighborhood-based models, using correlations, in predicting the bias of spent nuclear fuel decay heat calculations and identifying influential and highly similar benchmarks.

Coupled chemical–CFD modeling of unsealed dry storage of advanced test reactor spent fuel

Given the current lack of long-term disposal for spent nuclear fuel storage in the United States, the current inventory of advanced test reactor (ATR) fuel at Idaho National Laboratory (INL) is expected to be stored in a dry storage facility for an extended storage period. As part of an effort to understand the conditions during a 50-year period, a three-dimensional computational fluid dynamics (CFD) of the unsealed canisters packed with spent fuel in the storage facility was constructed, this CFD model is coupled with radiolytic chemical reactions associated with water vapor and other species. The coupled model is simulated over a range of sensitivity parameters for the fuel decay heat, relative humidity, and oxyhydroxide film thickness for undried and fully-dried fuel. The most important parameter was the initial fuel decay heat. Over the range of cases, the hydrogen and nitric acid concentrations are low enough no significant impact is expected.

Determination of the spent fuel decay heat with the VERA core simulator

VERA core simulator capabilities to predict spent fuel decay heat have been explored. The comparison with the Serpent2 and NEWT/TRITON sequence from the SCALE package, performed on the NPP Krško fuel test cases, has enabled independent verification of the successful implementation of the VERA depletion/decay calculation sequence applied on the extended nuclide set. Since the VERA is a 3-D core simulator, the focus in the rest of the paper was on the investigation of the 3-D effects and non-linearity of the soluble boron concentration averaging process, which are difficult to assess with most of available computational tools. A difference in the decay heat prediction between the 3-D and 2-D model approaches was analyzed. In addition, the impact of the Inconel grids on the decay heat was determined. Averaging the concentration of soluble boron over the fuel irradiation period has been found to be a viable approach, since it results in slight decay heat overprediction, which is considered conservative for most applications. Finally, demonstration on the Watts Bar Unit 1 cycle 1 depletion has shown formidable VERA capabilities to accurately predict spent fuel decay heat.

Sensitivity study of coupled chemical-CFD simulations for analyzing aluminum-clad spent nuclear fuel storage in sealed canisters

The United States Department of Energy (DOE) manages over 50 Metric Tons Heavy Metal (MTHM) of aluminum-clad spent nuclear fuel. One main source for DOE’s Aluminum-clad spent nuclear fuel (ASNF) inventory is the advanced test reactor (ATR) at the INL site, which makes this fuel of particular interest for storage scenarios. Road-ready and final disposition packaging configurations for the ATR fuel dictates storage within helium-backfilled, sealed DOE standard canisters. The conditions within these sealed canisters for extended (greater than50 year dry) storage is of interest. To further this goal, a three-dimensional (3D) multi-physics computational fluid dynamics (CFD) model is developed of the sealed DOE standard canisters. This 3D CFD model is one-way coupled with bulk gas radiolysis reactions considering sealed canisters with inert gas and possible trace amounts of air and water vapor. This study looks at the evolution of the thermal history of the canisters over a 50 year time period with a coupling to the chemical reactions occurring from radiolytic breakdown of residual water. A sensitivity study is then carried out over the parameters of the model including the fuel decay heat, residual water content, sealed pressure, canister external temperature, and canister emissivity. In pure helium, hydrogen generation rates are low, under 10 ppm, but hydrogen generation rates are affected greatly by the presence of even 1% residual air, increasing by 50-plus-fold, and nitric acid generation with residual air also occurs ranging from 500 to 4000 ppm after 50 years. The fuel decay heat and the residual water content show the most importance in the generation of hydrogen gas in pure air, and for nitric acid with a residual air condition. External temperature, canister emissivity and sealed pressure all show minor sensitivity effects to the generation of potentially harmful species.

Validation of spent nuclear fuel decay heat calculations using Polaris, ORIGEN and CASMO5

Decay heat calculations of spent nuclear fuel (SNF) using Polaris and ORIGEN codes in the SCALE code system, and CASMO5 code, are validated using measurements from the Clab and GE-Morris facilities. Multiple hypothesis testing, relying on permutations and bootstrapping, is conducted to simultaneously analyze the significance of differences between calculations and measurements – namely the biases. The biases are small for the Clab data, and larger and have more variance for the GE-Morris data. Then, statistical tests are applied separately based on the measurement laboratory and the SNF origin. The calculations on the Clab PWRs show to be systematically different from the measurements. The remaining benchmarks show to be insignificantly different from the measurements. The benchmarks are useful to validate decay heat calculations of PWR and BWR fuel assemblies at various cooling times. Nevertheless, the choice of the data and the codes could affect the significance of the obtained biases.

Optimisation of used nuclear fuel canister loading using a neural network and genetic algorithm

This paper presents an approach for the optimisation of geological disposal canister loadings, combining high resolution simulations of used nuclear fuel characteristics with an articial neural network and a genetic algorithm. The used nuclear fuels (produced in an open fuel cycle without reprocessing) considered in this work come from a Swiss Pressurised Water Reactor, taking into account their realistic lifetime in the reactor core and cooling periods, up to their disposal in the final geological repository. The case of 212 representative used nuclear fuel assemblies is analysed, assuming a loading of 4 fuel assemblies per canister, and optimizing two safety parameters: the fuel decay heat (DH) and the canister effective neutron multiplication factor k_{mathrm{eff}}. In the present approach, a neural network is trained as a surrogate model to evaluate the k_{mathrm{eff}} value to substitute the time-consuming-code Monte Carlo transport & depletion SERPENT for specific canister loading calculations. A genetic algorithm is then developed to optimise simultaneously the canister k_{mathrm{eff}} and DH values. The k_{mathrm{eff}} computed during the optimisation algorithm is using the previously developed artificial neural network. The optimisation algorithm allows (1) to minimize the number of canisters, given assumed limits for both DH and k_{mathrm{eff}} quantities and (2) to minimize DH and k_{mathrm{eff}} differences among canisters. This study represents a proof-of-principle of the neural network and genetic algorithm capabilities, and will be applied in the future to a larger number of cases.

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

DEVELOPMENT OF FUEL CHARACTERIZATION TOOL BASED ON LIBRARY INTERPOLATION

The feasibility of using Origen+ARP code for depletion and decay calculations for Krško NPP was tested by performing depletion and decay calculations using interpolated libraries and comparing the results to the ones calculated from non-interpolated libraries in order to evaluated the number of libraries needed in order to interpolate fuel properties with sufficient precision for any realistic burnup scenario. For Krško NPP fuel, using three interpolation libraries with different decay heat parameters was enough to bring the approximation error bellow 0:5 % when comparing fuel decay heat through the decay interval.

Validation of spent nuclear fuel decay heat calculation by a two-step method

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100–4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

Calculation of Evaporation From Fukushima Nuclear Power Plant Spent Fuel Pools

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and postaccident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The postaccident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25% lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

Effect of graphenenano-fluid on heat pipe thermal performance for passive heat removal in nuclear spent fuel storage pool

After Fukushima Dai-ichi nuclear power reactor accident, spent fuel in spent fuel storage pool (SFSP) became an important thing to pay attention to, due to its decay heat release. If station blackout occurred, the active system for SFSP cooling system will experience malfunction to remove the product of decay heat. To keep spent fuel safe, heat pipe as passive cooling system device can be used to remove spent fuel decay heat even if the active cooling system failed. Heat pipe with 6 m on length will be proposed as apassive cooling system in SFSP. For that, it is necessary to analyze the effect of Graphene nano-fluid as heat pipe working fluid. The objective of this research is to know the effect on Graphene nano-fluid to enhance the heat pipe thermal performance and to know the heat transfer phenomena inside the heat pipe based on Graphene nano-fluid. Graphene nano-fluid with 1% weight concentration was used as working fluid with filling ratio of 80%. The experimental investigation is conducted with varying the evaporator heat load of 1000, 1500, 2000, and 2500 W. Water as coolant flows in thecondenser with aconstant volumetric flow rate of 8 L/min. The experiment results show that there was anovershoot, zigzag and stable phenomena inside the heat pipe. The thermal resistance of heat pipe is obtained at 0.015 °C/W. The use of Graphene nano-fluid as working fluid can enhance the heat pipe thermal performance significantly and can be used as an alternative working fluid in heat pipe for thepassive cooling system in SFSP of nuclear power plant.

Conceptual design and analysis of heat pipe cooled silo cooling system for the transportable fluoride-salt-cooled high-temperature reactor

After the Fukushima Dai-ichi nuclear power plant severe accident, the advanced facilities that could improve the nuclear safety performance are preferred by the public. This paper discusses the preliminary design and thermal-hydraulic analysis of a heat pipe cooled passive heat removal system for the transportable fluoride-salt-cooled high-temperature reactor of 20MWth proposed by Massachusetts Institute of Technology in 2014. Under severe accident, the high temperature fuel and fluoride salt sustain at the reactor core, the heat pipes are inserted by gravity to remove the decay heat to the final heat sink of silo cooling system. To illustrate the feasibility of the designed passive heat removal system, the heat pipe startup performance is firstly numerically investigated by self-developed code HPTAC (Heat Pipe Transient Analysis Code). Four heat transfer limits are adopted as criteria for the success of heat pipe operation, 1) Viscous 2) Entrainment 3) Sonic 4) Capillary limits. The benchmark of heat pipe startup compared with experimental data is conducted as well as a sensitivity study for heat pipe inclination angle. Secondly, the transient performance of passive heat removal system based on reasonable assumptions of heat pipe model is analyzed. Results show that the heat pipe reaches the normal operation in 6min, in good agreement with experimental data with the maximum discrepancy of 16%. Most of fuel decay heat can be removed in 5h by heat pipe passive heat removal system and all the key temperature are kept below the allowed temperature limits.

COSI6: A Tool for Nuclear Transition Scenario Studies and Application to SFR Deployment Scenarios with Minor Actinide Transmutation

Nuclear systems, composed of reactors with varied fuel and cycle facilities (enrichment plant, fabrication plant, reprocessing plant, etc.), are complex and in constant evolution. Since 1985, the CEA has been developing the simulation software COSI to study different trajectories of nuclear fleet evolution and provide technical elements to decision makers. The principle of COSI, including the typical composition of the data set, is exposed. To evaluate as accurately as possible the isotopic compositions of materials, several physical models are implemented in COSI. The main ones, the evolution calculation and the equivalence models, are described in detail. An exercise of validation of COSI carried out on the French pressurized water reactor (PWR) historical nuclear fleet until 2010 is also presented, as well as a methodology for propagation of input uncertainties on COSI results.To illustrate the possibilities of COSI, the results of different scenarios studied in the framework of the French Act for Waste Management are discussed. The objective of these scenarios is to evaluate the feasibility of sodium-cooled fast reactor (SFR) deployment to renew the French PWR fleet on different timescales and to analyze the costs and the benefits of different options of minor actinide (MA) partitioning and transmutation. The impacts of SFR deployment on cycle facilities such as the fabrication plant, the spent fuel storage, and the reprocessing plant are minimized. The SFR deployment appears to be feasible with regard to fissile material availability, with an adaptation of fuel cooling time before reprocessing or of SFR breeding gain.Minor actinide transmutation in homogeneous mode MA diluted in core) and transmutation in heterogeneous mode (in MA-bearing blankets) are compared not only according to their impacts on cycle facilities and on ultimate waste but also according to the reduction of their inventories. The increases in fresh fuel thermal power and spent fuel decay heat due to the addition of MAs in fuels are quantified. The cases of transmutation of all MAs (americium, neptunium, and curium) and of americium only are distinguished. Alternative scenarios are explored to overcome the challenges associated with each option: reduction of the maximal MA content in fresh SFR fuel in the case of homogeneous transmutation and reduction of interim MA storage in the case of heterogeneous transmutation.

Analysis of AP1000 containment passive cooling system during a loss-of-coolant accident

A loss-of-coolant accident (LOCA) is one of the design basis accidents (DBAs) in nuclear power plant safety. DBAs may cause containment failure as a result of high temperature or pressure from LOCAs. This study investigates the integrity of the protective mechanism of the AP1000 containment system during a LOCA. The performance of the passive containment cooling system (PCCS) and the ability to perform decay heat removal for long-term cooling are evaluated in this study. The PCCS utilizes gravity-driven natural convection and naturally induced passive safety devices to move the decay heat of fuels to the atmosphere. In this study, two accidents are selected for analysis: a double-ended guillotine of a hot leg and a double-ended break of a main steam line. The analytical results are compared with the corresponding results provided in the AP1000 “Design Control Document” (DCD). The short-term calculations and comparisons with the DCD suggest that GOTHIC 7.2a with a natural convection model may properly represent the phenomenon of passive, safe heat removal. This paper also simulates the long-term calculations of spraying the outer primary containment with a water film and the steam condensation of the inner containment.

Validation of ORIGEN for LWR used fuel decay heat analysis with SCALE

The energy release rate from the decay of radionuclides can be a critical design parameter for used nuclear fuel storage, transportation, and repository engineered systems. Validation of the SCALE nuclear analysis code system capabilities in predicting decay heat for commercial used fuel applications has been performed using decay heat measurements for fuel assemblies irradiated in pressurized and boiling water reactors. The experimental data used for validation include a large number of full-length-assembly decay heat measurements that were performed between 2003 and 2010 at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, Clab, operated by the Swedish Nuclear Fuel and Waste Management Company, SKB. The measured fuel assemblies cover the burnup range 14–51GWd/MTU and cooling times between 12 and 27 years, which are times of interest to used fuel transportation and storage applications. The validation results indicate good agreement between calculated and measured decay heat values, generally within the reported measurement uncertainty. The effects of key modeling assumptions and data used in the calculations are presented and discussed.

Effect of minor actinides doping on plutonium produced in large-scale FBR blanket

The present study focuses on the effect of minor actinides (MAs) addition into the FBR blanket as ways of increasing fraction of even-mass-number plutonium isotopes, especially 238Pu, aiming at enhancing the proliferation resistance of plutonium produced in the blanket. The MA loading potential to enhance the proliferation resistance of plutonium is investigated, with considering actual design constraints on the fuel decay heat from the fuel handling and fabrication points of view, as MAs considerably generate decay heat. It reveals that depending on doping quantity of MAs, it is possible to denature produced plutonium by MA transmutation. MA addition in the blanket gives a significant increment in 238Pu fraction of generated plutonium but less effect on other even-mass-number plutonium isotopes. However, it is important that MA compositions should be adequately controlled to satisfy both the proliferation resistance requirements and the decay heat constraints for fuel handling.

Can Spent Nuclear Fuel Decay Heat Prevent Radionuclide Release?

AbstractHeat generated by radioactive decay of spent fuel represents a potentially important barrier to water accumulation on commercial spent nuclear fuel in breached waste packages. In the absence of water, fuel degradation and radionuclide release will be negligible. Thermal models for the proposed Yucca Mountain Repository suggest that, after a period of approximately 1000-4000 years (depending on loadingand ventilation conditions), the repository drift walls may decline to sub-boiling temperatures, thus allowing humidity in the drift to increase. The question thus arises, is the thermal gradient between the fuel and the drift sufficient to prevent water accumulation in a humid drift environment throughout the regulatory period? The answer depends on the balance between processes that oppose water condensation ontothe fuel (decay heat) and those that promote condensation such as the deliquescence of hygroscopic phaseswithin the fuel.Our experimental results indicate that deliquescence could lead to the condensation of water onto spent fuel despite the thermal “self-drying”effect if the following criteria are met: (1) the fission product salt CsI is present in the fuel or in the fuel-cladding gap, (2) the relative humidity in the driftexceeds 80% while temperatures in the waste package are around 90oC. Previous work suggests that these criteria may be met for some fuel pins within the proposed Yucca Mountain Repository. However,experiments that account for the role of U(VI) alteration phases suggest that deliquescence may be a self-limiting process in the sense that deliquescent components (e.g. Cs, Ba, Sr) may be incorporatedinto nondeliquescent U(VI) phases that form from the corrosion of spent fuel.