Release Of Volatile Fission Products Research Articles

Characterization of bremsstrahlung and γ-rays of fuel debris

The characterization of bremsstrahlung and γ-rays from fuel debris differs from that of spent fuels evaluated to date, due to factors such as material composition and release of volatile fission products. In this work, in order to clarify the conditions under which the effect of bremsstrahlung compared to the total photons (bremsstrahlung and γ-rays) in fuel debris is maximized, the average energies and dose rates from the energy spectra of bremsstrahlung and γ-rays on the fuel debris surface were obtained using a Monte Carlo simulation. In the simulation, the average energies and dose rates were evaluated with consideration of the composition, size, fission product release, and retrieval time of the fuel debris. The simulation showed that the composition with the largest amount of change to the average total photons energy caused by bremsstrahlung was the molten fuel debris, and the composition with the maximum fraction of bremsstrahlung in the dose rate was the UO2. The maximum value of the fraction of bremsstrahlung in the dose rate was evaluated to be about 17%. This work is expected to contribute to the prediction of the radiation characteristics of the fuel debris that will be retrieved from the Fukushima Daiichi Nuclear Power Station in the near future.

On the use of spectral algorithms for the prediction of short-lived volatile fission product release: Methodology for bounding numerical error

Recent developments on spectral diffusion algorithms, i.e., algorithms which exploit the projection of the solution on the eigenfunctions of the Laplacian operator, demonstrated their effective applicability in fast transient conditions. Nevertheless, the numerical error introduced by these algorithms, together with the uncertainties associated with model parameters, may impact the reliability of the predictions on short-lived volatile fission product release from nuclear fuel. In this work, we provide an upper bound on the numerical error introduced by the presented spectral diffusion algorithm, in both constant and time-varying conditions, depending on the number of modes and on the time discretization. The definition of this upper bound allows introducing a methodology to a priori bound the numerical error on short-lived volatile fission product retention.

A numerical methodology for estimation of volatile fission products release from nuclear fuel

In this paper we have developed a numerical methodology which is capable of estimating volatile fission product release from nuclear fuel under changing irradiation conditions with incorporation of all physical phenomena’s. The present study bridges the gap between previous analytical studies by providing estimation of unstable fission product release under accidental condition with and without incorporation of precursor’s history effect. Based on the present methodology a computer model “FIPRAP-Fission Product Release Analysis Program” has been developed and validated with standard fission product release problems and benchmarked against internationally accepted CORSOR-BOOTH model as well as Accident Source Term Evaluation Code (ASTEC). The result of the proposed model lies within 15% of internationally accepted models.

SARNET benchmark on Phébus FPT3 integral experiment on core degradation and fission product behaviour

The importance of computer simulations in the assessment of nuclear plant safety systems has increased dramatically during the last three decades. The systems of interest include existing or proposed systems that operate, for example, normal operation, in design basis accident conditions, and in severe accident scenario beyond the design basis. The role of computer simulations is especially critical if one is interested in the reliability, robustness, or safety of high consequence systems that cannot be physically tested in a fully representative environment. In the European 7th Framework SARNET project, European Commission (EC) co-funded from 2008 to 2013, the Phébus FPT3 experiment was chosen as a code benchmark exercise to assess the status of the various codes used for severe accident analyses in light water reactors.The aim of the benchmark was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor (PWR), starting from the initial stages of core degradation, fission product, actinide and structural material release, their transport through the primary circuit up to the behaviour of the released fission products in the containment.The FPT3 benchmark was well supported, with participation from 16 organisations in 11 countries, using 8 different codes. The temperature history of the fuel bundle and the total hydrogen production were well captured. No code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. The total volatile fission product release was well simulated, but the kinetics were generally overestimated. Concerning the modelling of semi-volatile, low-volatile and structural material release, the models need improvement, notably for Mo and Ru for which a substantial difference between bundle and fuel release was experimentally observed, due to retention in the cooler upper part of the bundle. The retention in the primary circuit was not well predicted, this was due mainly the non-prototypic formation of a boron-rich blockage in the rising line of the FPT3 steam generator, simulated in the circuit as a single external cooled U tube. The deposition mechanism and the volatility of some elements (Te, Cs, I) could be better predicted.Containment vessel thermal hydraulics, designed in the experiment to be well-mixed, were well calculated. Concerning the containment aerosol depletion rate, only stand-alone cases (in which the input data were derived from experimental data) provided acceptable results, whilst the integral cases (in which the input data came from circuit calculations) tended to largely overestimate the total aerosol airborne mass entering the containment.The disagreement of the calculated total aerosol airborne mass in the containment vessel with the measured one is due to the combination of a general underestimation of the overall circuit retention and overestimation of fission product and structural material release.Calculation of iodine chemistry in the containment turned out to be a major challenge. Its quality strongly depends on the correct prediction of chemistry speciation in the integral codes. The major difficulties are related to the presence of high fraction of iodine in gaseous form in the primary circuit during the test, which is not correctly reproduced by the codes. This inability of the codes compromised simulation of the observed iodine behaviour in the containment.In the benchmark a significant user effect was detected (different results being obtained by different users of the same code) which had to be taken into account in analysing the results. This article reports the benchmark results comparing the main parameters calculated and observed, summarising the results achieved, and identifying the areas in which understanding needs to be improved. Relevant experimental and theoretical work is under way to resolve the issues raised.

JRC's on-line fission gas release monitoring system in the high flux reactor Petten

For HTR fuel irradiation tests in the HFR Petten a specific installation was designed and installed, dubbed the “Sweep Loop Facility” (SLF). The SLF is tasked with three functions, namely temperature control by gas mixture technique, surveillance of safety parameters (temperature, pressure, radioactivity etc.) and analysis of fission gas release for three individual capsules in two separate experiments. The SLF enables continuous and independent surveillance of all gas circuits. The release of volatile fission products of the in-pile experiments is monitored by continuous gas purging. The fractional release of these fission products, defined as the ratio between release rate of a gaseous fission isotope (measured) to its instantaneous birth rate (calculated), is a licensing-relevant test for HTR fuel. The newly developed gamma spectrometry station allows for higher measurement frequencies, thus enabling follow-up of rapid and massive release transients. The designed stand-alone system was tested and fully used through the final irradiation period of the HFR-EU1 experiment which was terminated on 18 February 2010. Its robustness allowed us to use it as extra safety instrumentation. This paper describes the gas activity measurement technique based on HPGe gamma spectrometry and illustrates how qualitative and quantitative analysis of volatile fission products can be performed on-line.

A general model for predicting coolant activity behaviour for fuel-failure monitoring analysis

A mathematical treatment has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The fission product activity in both the fuel-to-sheath gap and primary heat transport system as a function of time can be predicted during all reactor operating conditions, including: startup, steady-state, shutdown, and bundle-shifting manoeuvres. In addition, an improved ability to predict the coolant activity of the 135Xe isotope in commercial reactors is discussed. A method is also proposed to estimate both the burnup and the amount of tramp uranium deposits in-core. The model has been validated against in-reactor experiments conducted with defective fuel elements containing natural and artificial failures at the Chalk River Laboratories. Lastly, the model has been benchmarked against a defective fuel occurrence in a commercial reactor.

Validation of Severe Accident Codes on the Phebus Fission Product Tests in the Framework of the PHEBEN-2 Project

Analyses of severe accidents in nuclear power plants by using integral codes are necessary in order to develop accident management strategies that prevent such accidents or mitigate their consequences for the environment. The most important requirement for the development of integral codes is to achieve good predictability of a given accident scenario through the understanding and quantification of severe accident phenomena and their underlying physical and chemical processes. In this paper, the progress in modeling the processes related to the radioactive source term, and in particular progress related to the release and transport of fission products in the circuit and containment, is demonstrated by the assessment of integral and detailed codes using the experimental results of the in-pile Phebus fission product tests (FPTs). It is shown that the integral codes are good in predicting both the hydrogen release and the total release of volatile fission products from the bundle.It is also shown that the commonly used fission product transport codes overestimate the deposited aerosol mass in the Phebus steam generator. However, by using an improved model for the thermophoretic aerosol particle deposition, it has been possible to reproduce the aerosol mass deposited in the steam generator more accurately. The containment analyses carried out with both lumped-parameter and multidimensional computational fluid dynamics codes showed that the measured thermal-hydraulic data are accurately reproduced. The aerosol behavior in the containment estimated from the lumped-parameter codes corresponded satisfactorily to the experimental data. The iodine chemistry codes highlighted the substantial role of silver released from the degraded absorber rod (Ag-In-Cd), as it was observed experimentally; however, the temporal dependence of the gaseous iodine concentration in the containment atmosphere was poorly calculated. There are plans to improve the modeling in order to reproduce better the fission product release from the bundle, the fission product transport in the primary circuit duct, and the gas phase chemistry in the containment, with particular emphasis on gaseous iodine species. Further plans include the analysis of Phebus FPT3, which was the last in the series of Phebus tests, with its boron-carbide control rod.

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO2 fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO2 to UO2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.

A model for predicting coolant activity behaviour for fuel-failure monitoring analysis

A mathematical treatment has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The activity in both the fuel-to-clad gap and coolant as a function of time can be predicted during all reactor operations including steady operation as well as reactor shutdown, startup and bundle-shifting maneuvers. The model has been implemented as the STAR (Steady-state and Transient Activity Release) code based on a finite-element solution of the mass transport equations. The model parameters are derived from in-reactor experiments conducted with defective fuel elements containing natural and artificial failures at the Chalk River Laboratories. The STAR code has also been successfully validated against an analytical solution and benchmarked against several defect occurrences in a commercial reactor.

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO2 fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO2 to UO2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.

Radionuclide Release from Mixed-Oxide Fuel under High Temperature at Elevated Pressure and Influence on Source Terms

The radionuclide release from mixed-oxide fuel (MOX) under severe accident conditions was investigated in the VEGA program to provide the technical bases for safety evaluation including probabilistic safety assessment (PSA) for light water reactor (LWR) using MOX. The MOX specimens irradiated at Advanced Thermal Reactor (ATR) Fugen were heated up to 3,123K in helium at 0.1 and 1.0MPa. The release of volatile fission products (FP) was slightly enhanced below 1,623 K compared with that of UO2. The volatile FP release at elevated pressure was decreased as in the case with UO2. The total fractional release of Cs reached almost 100% while almost no release of low-volatile FP even after the fuel melting. The release rate of plutonium above 2,800 K increased rapidly although the amount was small. Since the existing models cannot predict this increase, an empirical model was prepared based on the data. The present study showed that there are no large differences in total fractional releases and inventories of important FP in PSA between UO2 and MOX. This suggests that the consequences of LWR using MOX are mostly equal to those using UO2 from a view point of risks.

Thematic network for a Phebus FPT1 international standard problem (THENPHEBISP)

The THENPHEBISP 2-year thematic network started in December 2001, and was concerned with OECD/CSNI International Standard Problem 46, itself based on the Phebus FPT1 core degradation/source term experiment. The aim was to assess the capability of computer codes to model in an integrated way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. ISP-46, coordinated by IRSN/DRS Cadarache, attracted 33 participating organisations, from 23 countries and international bodies, who submitted 47 base case calculations and 21 best-estimate calculations, using 15 different codes. The thermal behaviour of the fuel bundle and the hydrogen production were generally well captured, and good agreement for the core final state could be obtained with a suitable choice of bulk fuel relocation temperature, however this is unlikely to be representative of all plant studies so sensitivity calculations are needed with the modelling in its current state. Total volatile fission product release was simulated, but its kinetics, and the overall modelling of semi-volatile, low-volatile and structural material release (Ag/In/Cd, Sn) needs improvement. Overall retention in the circuit is well predicted, but calculations underestimate deposits in the upper plenum and overestimate those in the steam generator, also the volatility of some elements could be better predicted. Containment thermal hydraulics and depletion rate of aerosols are well calculated, but with difficulties related to partition amongst the deposition mechanisms. Calculation of iodine chemistry in the containment turned out to be more difficult. Its quality strongly depends of the calculation of release and transport in the integral codes. The major difficulties are related to the existence of gaseous iodine in the primary circuit and to the prediction of the amount of organic iodine in the gas phase. This paper summarises the results achieved and the implications for plant calculations.

Radionuclide Release from Mixed-Oxide Fuel under High Temperature at Elevated Pressure and Influence on Source Terms

The radionuclide release from mixed-oxide fuel (MOX) under severe accident conditions was investigated in the VEGA program to provide the technical bases for safety evaluation including probabilistic safety assessment (PSA) for light water reactor (LWR) using MOX. The MOX specimens irradiated at Advanced Thermal Reactor (ATR) Fugen were heated up to 3,123K in helium at 0.1 and 1.0MPa. The release of volatile fission products (FP) was slightly enhanced below 1,623 K compared with that of UO2. The volatile FP release at elevated pressure was decreased as in the case with UO2. The total fractional release of Cs reached almost 100% while almost no release of low-volatile FP even after the fuel melting. The release rate of plutonium above 2,800 K increased rapidly although the amount was small. Since the existing models cannot predict this increase, an empirical model was prepared based on the data. The present study showed that there are no large differences in total fractional releases and inventories of important FP in PSA between UO2 and MOX. This suggests that the consequences of LWR using MOX are mostly equal to those using UO2 from a view point of risks.

Artificial neural network models for volatile fission product release during severe accident conditions

Artificial neural network (ANN) models have been developed to predict the release of volatile fission products from both Canada deuterium uranium (CANDU) and light water reactor (LWR) fuel under severe accident conditions. The CANDU model was based on data for the release of 134Cs measured during three annealing experiments (Hot Cell Experiments 1 and 2, or HCE-1, HCE-2 and metallurgical cell experiment 1, or MCE-1) at Chalk River Laboratories. These experiments were comprised of a total of 30 separate tests. The ANN established a correlation among 14 separate input variables and predicted the cumulative fractional release for a set of 386 data points drawn from 29 tests to a normalized error, E n, of 0.104 and an average absolute error, E abs, of 0.064. Predictions for a blind validation set (test HCE2-CM6) had an E n of 0.064 and an E abs of 0.054. From this 14 variable ANN model, a pruned version utilizing only the 6 most significant variables was trained to provide comparable predictions. An ANN model was also developed for LWR fuel, based on data from the vertical induction (VI) series of tests (VI-2 to VI-5) conducted at Oak Ridge National Laboratory. Predictions for data not used in ANN training had an E n of 0.045 and an E abs of 0.059. A methodology is presented for deploying the ANN models by providing the algorithms for trained ANNs and the corresponding connection weights. Finally, the performance of the full ANN CANDU model was compared to a fuel oxidation model developed by Lewis et al. and to the US Nuclear Regulatory Commission's CORSOR-M.

Low volatile fission-product release and fuel volatilization during severe reactor accident conditions

An analytical model has been developed to describe the release behavior of low-volatile fission products from uranium dioxide fuel under severe reactor accident conditions. The effect of the oxygen potential on the chemical form and volatility of fission products is determined by Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix or fission product vaporization from the fuel surface. The effect of fuel volatilization (i.e., matrix stripping) on the release behavior is also considered. The model has been validated against several out-of-pile annealing experiments performed at high temperature in various oxidizing and reducing conditions.

Dissolution of UO 2 in molten Zircaloy-4 Part 1: Solubility from 2000 to 2200°C

This paper reports solubility measurements for unirradiated UO 2 fuel at 2000, 2100 and 2200°C in initially O-free Zircaloy-4 and at 2100 and 2200°C in Zircaloy-4 with an initial 25 at% O content. The Zircaloy/25% O was used to represent the O-saturated Zircaloy component of steam-oxidized cladding. A UO 2/Zircaloy mass ratio of ~ 10.9 was used in most experiments to simulate the quantities present in a CANDU-PHW ( CAN ada D euterium U ranium P ressurized Heavy Water ) reactor fuel bundle. At each temperature, the UO 2 solubility reached a maximum with initially O-free Zircaloy, indicating that fuel dissolution in previously unoxidized cladding represents a worst-case scenario during a severe fuel damge (SFD) accident. Thus, use of UO 2 solubilities to model SFD accidents allows upper limits to be placed on possible release of volatile fission products at any one temperature and may be more defensible than the use of kinetic data for modeling fuel dissolution.

Modeling of Fission Product Release and Transport for Severe Fuel Damage Analyses

In this paper best-estimate calculations of realistic source terms are presented that reduce uncertainties in predicting volatile fission product release from the UO{sub 2} fuel over the temperature range from 770 to 2500 K. The proposed method of correlation includes such fuel morphology effects as equiaxed fuel grain growth and fuel/cladding interaction. The method correlates the product of fuel release rate and equiaxed grain size with the inverse fuel temperature to yield a bulk mass transfer correlation. It is found that fewer and slower releases are predicted utilizing the bulk mass transfer correlation than with the steam oxidation model and the U.S. Nuclear Regulatory Commission's NUREG-0956 correlation. Computational modules are developed to perform the thermal-hydraulic and fission product calculations needed to analyze the severe fuel damage tests. The predictions utilizing the bulk mass transfer correlations overall follow the experimental time-release histories during the heatup, power hold, and cooldown phases of the transients. Good agreements are achieved for the integral releases both in timing and in magnitude. The proposed bulk mass transfer correlations can be applied to both current and advanced light water reactor fuels.

Analytical solution of the radioactive volatile fission product release equation in the presence of precipitation and re-solution

The diffusion transport equation of volatile fission products in nuclear fuels in the presence of precipitation, re-solution and radioactive decay was analytically solved for stationary and non-stationary conditions. Comparisons are made of exact solutions with those preliminarily obtained with formally simplified methods; they show excellent agreement.

Analysis of Fission Product Release Behavior from the Three Mile Island Unit 2 Core

AbstractAn analysis of fission product release during the Three Mile Island Unit 2 (TMI-2) accident has been performed to provide an understanding of fission product behavior that is consistent with both the best-estimate accident scenario and fission product results from the sample acquisition and examination efforts. “Firstprinciples” fission product release models are used to describe release from intact, disrupted, and molten fuel during the various phases of the TMI-2 accident.Extensive gaseous and volatile fission product release is calculated to have occurred, with local regions of the core experiencing up to 100% release. Diffusion is calculated to have dominated release during the initial core heatup, while bubble coalescence and rise dominated release from the large consolidated region of molten material. The calculations are generally consistent with fission product retention data from upper and lower plenum debris bed samples. An exception to this is the small retention of cesium in the lower ...

Oxide fuel transients

The behavior of UO 2 fuel from power reactors has been studied up to high burn-up (~ 60 GWd/tU), under both steady state and power transient conditions in the range of 35 to 48 kW/m. Careful post-irradiation examination involving electron probe microanalysis, transmission and scanning electron microscopy in addition to detailed hot cell examination have provided a large data base on the radial migration and release of volatile fission products. Theoretical evaluation with different computer codes and supporting laboratory experiments established the basis for understanding the kinetics and mechanisms operative during transients in high burn-up UO 2. Typical examples are given to demonstrate the degree of understanding achieved