Fission-product measurements in the SCARABEE mono and seven-pin loss-of-coolant experiments

During the course of the ‘SCARABEE’ experimental programme, the primary objective of which was to study the fuel and thermohydraulic behaviour, fission-product concentration were measured by a number of techniques. These included delayed neutron monitoring, total gamma, gamma spectrometry, sodium sampling and analysis and gas sampling and analysis. This paper describes the methods used to evaluate the data from these measurements and discusses the results.The theoretical models for fission-product generation and release are described and compared with the experimental data, and a critical examination of the techniques and results leads to recommendations for improvements in methods for future experiments. Conclusions are drawn relating to fission-product behaviour in LMFBRs. It is recognized that the experiments do not model closely the power reactor situation, but it is argued that the results obtained, particularly from delayed-neutron measurements, do provide a useful pointer to the value of fission-product monitoring in limiting fuel damage from local fuel overheating faults.     

On the nature of aerosols produced during a severe accident of a water-cooled nuclear reactor

Particle behaviour depends strongly on classic characteristics, e.g., size, and less macroscopic ones involving structure and composition these being especially important in situations of strong differential forces on a particle, i.e., surface impact or intensely-shearing flows. The former situation may lead to particle deposition or break-up and re-entrainment (with potential accident-management implications). This paper reviews information on aerosols from prototypical experiments identifying common features and typical variations. It emerges that a particle comprising one-third metal, one-third metal oxide and one-third a mixture of fission-product species would not be out of place in any potential reactor-accident sequence. Particle shapes appear relatively compact without branching chain-like structures. On size and structure, aerosols in the upstream part of the primary circuit would comprise a near-lognormal population with AMMD no more than 2 μm and geometric standard deviation around 2, particles comprising agglomerates of highly-coordinated clusters as small as 0.1 μm. In the containment, aerosols can typically be represented by primary-circuit particles and their agglomerates though particular circumstances (core–concrete interaction, hot-leg accident sequence) can alter this simple picture.     

Interpretation of fission-product transport behaviour in the Phébus FPT0 and FPT1 tests

Phébus FP studies the phenomenology of severe accidents in water-cooled nuclear reactors. Tests cover fuel rod degradation and behaviour of fission-products released via the coolant system into the containment. Analysis using computer codes aims to identify modelling weaknesses. Regarding fission-product behaviour in the coolant circuit, analyses of tests FPT0 and FPT1 are presented that used a standard version of a code, SOPHAEROS, with input data based solely on measured boundary conditions. Disagreements between calculated and experimental results are explored and plausible explanations presented. It is shown that in laminar conditions for a supersaturated vapour with competing homogeneous nucleation, heterogeneous nucleation and condensation on structures, codes can significantly underestimate the latter if entrance effects are ignored. Where thermophoresis dominates in hydrodynamically developed, weakly turbulent flow, codes can overestimate deposition; the likely explanation is underestimated mechanical resuspension. Concerning the transport of vapour species, it is shown that observations are compatible with passage of non-negligible amounts of the gas hydrogen iodide through the circuit. The final aspect of the exercise concerns deposit remobilization where this was different in the two tests and the understanding of which remains more speculative. Explanation invokes vibration of the apparatus producing mechanical resuspension in FPT0 and steam reacting with caesium deposits producing caesium hydroxide in FPT1.     

The chemical effects of composition changes in irradiated oxide fuel materials II-Fission product segregation and chemical equilibria

Four miniature fuel pins containing uranium/plutonium oxide, covering a wide range of oxygen-to-metal ratios, have been irradiated in a thermal reactor. These pins were subjected to a detailed examination, by electron probe microanalysis and ceramography. The results of this analysis have been interpreted in the light of the results of a computer model of the pin behaviour. The widely different fuel chemistry in each of the fuel pins produced different fission-product phases and different clad corrosion phenomena. The phases found are in general agreement with the model which predicts increasing oxygen-to-metal ratio with increasing burn-up and oxygen redistribution in the temperature gradient. Clad corrosion is seen to be inhibited at low oxygen potentials but the distinction between intergranular and ‘broad front’ corrosion is not attributable simply to the local oxygen potential.     

Primary coolant chemistry in VVER units

Operation experience with VVER coolant chemistry has been reviewed. The paper describes the results of measurements of radiation fields in primary system components and occupational doses that are compared with radiation control philosophy based on boron-potassium-ammonia coordinated water chemistry, modified and hydrazine water chemistry. The difference in water chemistry guidelines between VVERs and different current operational practices at VVER utilities are outlined. Special emphasis is given to the ammonia-hydrazine water chemistry on some VVER plants, high temperature filtration and silver behaviour.     

Advancement of modeling deposition and coagulation of aerosols in a nuclear reactor

The objective of the paper is to advance a code for calculating the behaviour of aerosol-shaped fission products (FP) in the primary circuit of a nuclear reactor. This FP may come into the circuit from the fuel during a severe accident. The paper is focused on development of models for predicting the deposition and coagulation rates of aerosols. Comparisons of model predictions with experimental data and direct numerical simulations are discussed.     

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.     

Towards a better understanding of iodine chemistry in RCS of nuclear reactors

The Phebus FP in-reactor integral experiments provided new insights into iodine transport through the primary circuit. Indeed, in these tests transported iodine was often found not associated with caesium as generally postulated up to now. Several iodine species were experimentally shown to have been transported in the hot leg at 700 °C, while a fraction was also suspected to be in a gaseous form in the cold leg at 150 °C. For a better estimate of the iodine source term to the containment, both in terms of speciation and quantity, it becomes thus necessary to reconsider iodine species behaviour along their pathway in the reactor coolant system (RCS).This paper presents the current understanding, mainly based on SOPHAEROS equilibrium chemistry calculations of Phebus FP tests performed within the I-RCS technical circle of the SARNET network of excellence in the EU 6th Framework programme. The suspected connection existing between Cs, Mo, Cd and I chemistry and the strong influence of both their release kinetics and related species thermodynamic properties on the iodine speciation in different environments (reducing/oxidizing) are highlighted. Potential explanations for the predicted iodine volatility and the level of association of I to Cs are also discussed.     

Deterministic Assessment for Environmental Impact of Yucca Mountain Repository Measured by Radiotoxicity

Environmental impact of the Yucca Mountain Repository (YMR) has been quantitatively and deterministically evaluated in terms of the radiotoxicity of transuranic (TRU) and fission-product radionuclides existing in the environment after released from failed packages. Inventory abstraction has been made based on the data published in Final Environmental Impact Statement by US Department of Energy (DOE). Mathematical model and computation code have been developed based on analytical solutions. Environmental impact from the commercial spent nuclear fuel (CSNF) packages is about 90% of the total impact including the contribution from defense waste (DW) packages. Impacts due to isotopes of Cm, Am, Pu and Np, and their decay daughters are dominant, compared with those from fission-product nuclides. Numerical results show that reduction of the TRU nuclides by a factor of 100 makes the impact from CSNF smaller than that from DW.     

Energy Dependence of Plutonium Fission-Product Yields

A method is developed for interpolating between and/or extrapolating from two pre-neutron-emission first-chance mass-asymmetric fission-product yield curves. Measured ²⁴⁰Pu spontaneous fission and thermal-neutron-induced fission of ²³⁹Pu fission-product yields (FPY) are extrapolated to give predictions for the energy dependence of the n + ²³⁹Pu FPY for incident neutron energies from 0 to 16 MeV. After the inclusion of corrections associated with mass-symmetric fission, prompt-neutron emission, and multi-chance fission, model calculated FPY are compared to data and the ENDF/B-VII.1 evaluation. The ability of the model to reproduce the energy dependence of the ENDF/B-VII.1 evaluation suggests that plutonium fission mass distributions are not locked in near the fission barrier region, but are instead determined by the temperature and nuclear potential-energy surface at larger deformation.     

Generation and validation of ORIGEN-S libraries for depletion and transmutation calculations based on JEF2.2 and EAF3 basic data

The data libraries for light elements, actinides and fission products of the ORIGEN-S code for depletion and transmutation calculations in the SCALE4.1 computer code system have been updated with respect to cross-section data, radioactive-decay data and fission-product yield data using JEF2.2 as the basic data source and EAF3 as an additional source. This required the fission-product library to be extended with 201 new fission-product nuclides or isomeric states. The effect of the update of different quantities involved is evaluated with a burn-up benchmark. When ORIGEN-S is used as a stand-alone code, i.e. without regular update of cross-sections of the major nuclides due to changes in the neutron spectrum during burn-up, the results show appreciable differences in actinide and fission-product densities due to the cross-section update. The effects of updates of decay data and fission-product yields are generally small, but with noticeable exceptions. The update of fission and capture reaction energies gives a small but systematic change in actinide and fission-product concentration. The new ORIGEN-S libraries have also been converted for use with the SCALE4.2 package.     

SARNET integrated European Severe Accident Research—Conclusions in the source term area

The overall aim of the SARNET (Severe Accident Research NETwork), in the EU 6th Framework programme was to integrate in a sustainable manner the research capabilities of fifty-one European organisations from eighteen member states of the European Union (EU) plus the Joint Research Centres, with one Canadian company, to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasised integrating activities, spreading of excellence (including knowledge transfer) and jointly executed research, with the knowledge gained being encapsulated in the European severe accident modelling code ASTEC. This paper summarises the achievements over the whole project in the Source Term Topic, which dealt with potential radioactive release to the environment, covering release of fission products and structural materials from the core, their transport in the primary circuit, and their behaviour in the containment.The main technical areas covered, as emphasised by the earlier EURSAFE project, were the effect of oxidative conditions on fission product release and transport (especially the behaviour of the highly radiotoxic ruthenium under air ingress conditions), iodine volatility in the primary circuit, control rod aerosol release (Ag-In-Cd) that affects iodine transport, containment by-pass in the case of steam generator tube rupture, aerosol retention in containment cracks, aerosol remobilisation in the circuit, and iodine/ruthenium behaviour in the containment especially concerning the volatile fraction in the atmosphere. The studies also covered performance of new experiments, analysis of existing data, and formulation and improvement of theoretical models. Significant progress was made in each area. Looking to the future, the 7th Framework successor project SARNET2 covers the remaining issues concerning iodine and ruthenium, including practical application of the results. The results outlined here will make a good basis for this continued endeavour.     

Containment analysis on the PHEBUS FPT-0, FPT-1 and FPT-2 experiments

In a severe accident, most of the fission-product species are already condensed in aerosols when they are released to the containment. The behaviour of these aerosol particles controls the fission-product transport into the containment and affects the global Source Term. The calculations presented here were performed using the CPA module (Containment Package implemented in the European integral code ASTEC) for the in-pile PHEBUS FPT-0, FPT-1 and FPT-2 experiments and are focused on the aerosol transport. A detailed thermal-hydraulic model was used in the CPA/ASTEC code to evaluate the gas circulation pattern in the closed containment volume. The comparison of ASTEC results showed that the patterns are similar to the ones predicted by the CFD-based codes. Good agreement was reached with the measured average thermo-hydraulic parameters such as containment gas pressure, temperature and the condensation rate on the condensers. The calculations with the detailed simulation of the flow in the PHEBUS containment qualitatively predicted the particle settling on the elliptic bottom and deposition on the painted wet condenser surfaces. It was shown that the influence of the gas circulation leads to a relatively quick mixing of aerosols in the containment atmosphere. In the tests investigated, the effect of the gas circulation on the airborne aerosol mass during the aerosol injection period is small because the injected mass flux is significantly higher compared to the deposition fluxes on the vessel surfaces. During the long-term aerosol deposition phase, the flow fields predicted by CPA/ASTEC have a medium impact on the evolution of the airborne mass in the PHEBUS containment.     

Irradiation performance of experimental fuel particles coated with silicon-alloyed pyrocarbon: A review

A review has been conducted on the use of silicon-alloyed pyrocarbon (Si-PyC) as an improved coating material for the two types of fuel particles used in the cores of high-temperature gas-cooled reactors. Based on recent data from extensive irradiation testing and postirradiation annealing of such experimental fuel particles, it is concluded that Si-PyC coatings offer considerable promise as replacements for the standard pure pyrocarbon (PyC) coatings used on thorium-based fertile fuels that have BISO coating designs. The primary advantage here is improved retention of fission products from bred U-233, with diffusion coefficients being as much as 100 times smaller for Si-PyC than for PyC. However, there is no significant improvement in mechanical performance of Si-PyC coatings over standard PyC coatings under irradiation. As a result, there is no incentive for using these coatings on TRISO particle designs of the type used on uranium-based fissile fuels, because here a silicon carbide barrier layer provides superior fission-product retention.     

Fission-product behaviour in irradiated TRISO-coated particles: Results of the HFR-EU1bis experiment and their interpretation

It is important to understand fission-product (FP) and kernel micro-structure evolution in TRISO-coated fuel particles. FP behaviour, while central to severe-accident evaluation, impacts: evolution of the kernel oxygen potential governing in turn carbon oxidation (amoeba effect and pressurization); particle pressurization through fission-gas release from the kernel; and coating mechanical resistance via reaction with some FPs (Pd, Cs, Sr). The HFR-Eu1bis experiment irradiated five HTR fuel pebbles containing TRISO-coated UO₂ particles and went beyond current HTR specifications (e.g., central temperature of 1523 K). This study presents ceramographic and EPMA examinations of irradiated urania kernels and coatings. Significant evolutions of the kernel (grain structure, porosity, metallic-inclusion size, intergranular bubbles) as a function of temperature are shown. Results concerning FP migration are presented, e.g., significant xenon, caesium and palladium release from the kernel, molybdenum and ruthenium mainly present in metallic precipitates. The observed FP and micro-structural evolutions are interpreted and explanations proposed. The effect of high flux rate and high temperature on fission-gas behaviour, grain-size evolution and kernel swelling is discussed. Furthermore, Cs, Mo and Zr behaviour is interpreted in connection with oxygen-potential. This paper shows that combining state-of-the-art post-irradiation examination and state-of-the-art modelling fundamentally improves understanding of HTR fuel behaviour.     

The ratio of grain boundary energy to surface energy of nuclear ceramics as determined from pore geometries

Several workers have derived high grain-boundary to surface energy ratios from measurement of the geometries of pores located at grain boundaries. The values (of more than unity) are two to three times greater than those calculated from the geometries of interactions of free surfaces with grain boundaries, and, in the case of irradiated UO₂, have been attributed to fission-product segregation effects. This work has confirmed previous observations of pore geometries and led to similar data for unirradiated and irradiated nuclear ceramics. It has been found that irradiation has no clear effect on pore geometries and hence it is concluded that fission-product segregation is not a significant factor. The work suggests that the theories used to derive energy ratios from pore geometries need modification.     

Seismic analysis of PWR-RCC fuel assemblies

The present paper investigates the dynamic behaviour of PWR-RCC fuel assemblies under seismic excitation. A simple vibrational model of the fuel assembly is proposed, which leads to natural frequencies whose spacing agree with experimental data. Available experimental results are reviewed. Impact characteristics of Zircaloy spacer grids are also discussed. It is proposed that their soundness criteria be expressed in terms of impact energy rather than in terms of impact force. The computer code CLASH is briefly described; it is utilized to perform a sensitivity analysis. An example of application is also given.     

Fracture mechanics investigations on cylindrical large scale specimens under thermal shock loading

To investigate the crack growth and crack arrest behaviour of primary circuit materials large scale experiments were conducted on component-like specimens under pressurized thermal shock loading at MPA Stuttgart. The material characteristics varied from high tough material to low tough material with higher nil ductility transition temperature to simulate EOL or beyond EOL-state. All tests started from in-service conditions and were cooled down to room temperature. The specimens showed both stable and unstable crack growth and partly crack arrest. The crack growth behaviour was verified by post test calculations and could be explained with the help of the multiaxiality of the stress state.     

Status of delayed-neutron precursor data: half-lives and neutron emission probabilities

We present in this paper a compilation of the present status of experimental delayed-neutron precursor data; i.e. β-decay half-lives (T_1/2) and neutron emission probabilities (P_n) in the fission-product region (27 ≤ Z ≤ 57). These data are compared to two model predictions of substantially different sophistication: (i) an update of the empirical Kratz-Herrmann formula (KHF), and (ii) a unified macroscopic-microscopic model within the quasi-particle random-phase approximation (QRPA). Both models are also used to calculate so far unknown T_1/2 and P_n values up to Z = 63. A number of possible refinements in the microscopic calculations are suggested to further improve the nuclear-physics foundation of these data for reactor and astrophysical applications.