Fission Product Source Research Articles

Computational evaluations and regulation of radioactive fission product sources during analysis of radiological consequences for active and planned nuclear power plants with VVER

The principles and approaches to evaluating a source of fission products when analyzing radiological consequences are described. Some of them were developed by adjusting previously used approaches. The most important and fundamental of these adjustments is an approach to regulating the number of leaky fuel elements for normal operation regimes in the calculation of the operational limit on the activity of fission products in coolant, for which the maximum admissible radioactive emissions and the irradiation dose due to them to the population living near a plant are not exceeded.

Secondary neutrons as the main source of neutron-rich fission products in the bombardment of a thick U target by 1 GeV protons

The diffusion-effusion model has been used to analyse the release and yields of Fr and Cs isotopes from uranium carbide targets of very different thicknesses (6.3 and 148 g/cm2) bombarded by a 1 GeV proton beam. Release curves of several isotopes of the same element and production efficiency versus decay half-life are well fitted with the same set of parameters. Comparison of efficiencies for neutron-rich and neutron-deficient Cs isotopes enables separation of the contributions from the primary (p + 238U) and secondary (n + 238U) reactions to the production of neutron-rich Cs isotopes. A rather simple calculation of the neutron contribution describes these data fairly well. The FLUKA code describes the primary and secondary-reaction contributions to the Cs isotopes production efficiencies for different targets quite well.

Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA). The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1 m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

Absorption of optical radiation by nanoclusters of sputtered metallic uranium and its oxides

Thin films of uranium-containing materials are used as energy-releasing elements in nuclear-pumped lasers. They serve as a source of fission products which ionize and excite the gaseous medium. In this pumping method, a uranium layer is sputtered by its own fission products. This is one of the factors that limit the service life of energy-releasing elements. The products of sputtering can enter the gas and affect the optical and kinetic properties of laser media. In the present work, calculations are performed of the electrodynamic cross-sections for photoabsorption of optical radiation by nanoparticles of metallic uranium and its oxides. The main space-time relaxation parameters are determined. The effect of sputtered uranium-containing nanoparticles on the absorption of optical radiation in a laser element is evaluated.

Testing hot cell shielding in the fuel conditioning facility.

A comprehensive shield test program for a hot cell complex, the Fuel Conditioning Facility at Argonne National Laboratory, has been completed with minimum radiation exposure to participants. The recently modified shielding design for two hot cells and their associated transfer paths for irradiated materials was analyzed and tested for attenuating gamma rays from mixed fission product sources. Testing was accomplished using 0.37 TBq (10 Ci) and 518 TBq (14,000 Ci) 60Co sources. Of specific concern were radiation levels around wall penetrations and the interface between transport casks and the cell floor. Detailed measurements were made for surfaces that bound the hot cells, a transfer tunnel between the two cells, and storage pits that extend below the floors of both cells. In addition to surface measurements, dose equivalent rates in adjacent corridors were determined when the larger source was exposed. Results indicate that with some administrative controls, the facility shields are adequate to meet the design criterion that limits annual dose to less than 10 mSv (1 rem) for facility workers.

Experiments and analysis of fission product release in HEU-fuelled SLOWPOKE-2 reactors

Fission product activity levels have been measured using a transportable gamma ray spectroscopy system at four SLOWPOKE-2 facilities. Through an analysis of the concentrations of these radionuclides in samples of the reactor coolant and gas headspace, the rate of release from the fuel has been determined by a Savitzky-Golay method and also by a non-linear least squares method. The release rate calculation has been validated against the mainframe code SUMRT. By examining the release rates, the source of the short-lived fission products is determined to be direct recoil from exposed uranium-bearing surfaces.

Volatile fission product and sodium release from liquids

For the improvement of radioactive source term calculations the computer code revols has been developed for the mechanistic modeling of the evaporative release of volatible species (e.g. water, sodium and volatile fission products as NaJ, Cs and Rb) from different hosts into an inert gas atmosphere. The code, showing a modular structure, has been developed to be coupled with reactor containment safety analysis codes as the contain / lmr and lmfbr version. In substituting existing constant-retention-factor formulations by introducing a geometry and state dependent, instationary retention factor, an improved aerosol and fission product source calculation can be obtained. The comparison of theoretical predictions with experimental results performed at the Karlsruhe Research Center shows good agreement.

Uncertainties remaining in severe accident phenomena and related research activities at JAERI

Abstract In utilizing PSAs to safety-related issues such as provision of accident management, it is essentially important to understand the uncertainties in the results. However, large uncertainties still exist in level 2 PSA results even though worldwide research programs have been carried out to understand and analyze severe accidents. This paper summarizes the uncertainties due to lack of knowledge on severe accident phenomena and discusses their effects on the estimated severe accident progression and fission product source terms, mostly based on the researches conducted at the Japan Atomic Energy Research Institute (JAERI). Furthermore, the paper also discusses the uncertainties caused by incompleteness and simplification in modeling. Since most of the uncertainties in physical models will not be significantly reduced in the near future, the methodology is expected to effectively utilize the PSA results in decision making even with such large uncertainties, while continuous research efforts to reduce the uncertainties are required.

Fabrication and Leach Testing of Synroc Containing Actinides and Fission Products

ABSTRACTSynroc containing mixed fission products and four actinides (237Np, 239Pu, 241Am and 244Cm) was fabricated by hot pressing on a 75 g-scale in separate hot-cell and glove-box facilities. Samples were physically characterised by density, which ranged from 98.2 to 99.1% of theoretical, by scanning electron microscopy (SEM) and by the cc-track etch technique to determine the distribution of actinide elements.The chemical durability of specimens prepared from the hot-pressed Synroc was determined in 460 leach tests carried out under static conditions at 70°C. Release of 137Cs dominated the fission-product source term while 237Np was the most leachable actinide element. Short-term release is dominated by the initial spilce from Cs and other readily soluble species at grain boundaries and in non-equilibrium phases. Long-term release is probably controlled by matrix solubility. A phenomenological model is developed which accounts for both short- and long-term release.

Evaluation of uncertainties in risk emanating from a number of severe accident issues

A study has been performed to estimate, for a particular pressurized water reactor, the uncertainty in risk associated with a number of key phenomenological issues. A second objective was to distinguish the individual importance of the various issues as contributors to the overall uncertainty in risk. The issues considered touched upon the areas of system behavior, containment loading, containment performance, and fission product source term behavior. It was found that the most important source of uncertainty for the plant in question (Surry) was direct containment heating (i.e., the transfer of heat from the core debris to the containment atmosphere when the debris is ejected at high pressure from the reactor vessel and dispersed throughout the atmosphere). Other significant issues included hydrogen burning, containment failure pressure, aerosol agglomeration uncertainties, the frequency of check valve failures leading to a loss-of-coolant accident (LOCA) outside containment, and the potential for having a LOCA induced by high temperatures in the reactor coolant system.

The preparation and characterization of a Ni-3 wt.% U reference fission product source for super phenix

The fabrication and quality control of Ni-3 wt.% U (93% 235U) alloy rods (diameter: 8 mm, total length: 8 m) used as a reference fission product source (RFPS) are described. This RFPs is needed for the calibration of the fuel element rupture detector which is used for the continuous monitoring of the delayed neutron activity in the primary sodium of the Super Phenix fast neutron reactor. Special care was taken to obtain a sufficient alloy purity, a bright surface finish and a homogeneous 235U content on the surface of the rods. Finally the usefulness of the RFPS in Super Phenix has been demonstrated.

The determination of uranium in Ni-3 wt.% U alloys by selective precipitation and subsequent α-counting

Ni-3 wt.% U alloys enriched at 93% in 235U are used as reference fission product source (RFPS) for the calibration of fuel element rupture detectors. One of the requirements is that 235U present in a surface layer of 10 μm thick is homogeneous at ±10%. A method was developed to control this homogeneity and certify the mean value of the uranium mass friction. It is based on selective precipitation of uranium and subsequent α-counting. Under optimal conditions accuracies of the order of 1% (1 s) are obtained.

Vapor Transport of Fission Products in Postulated Severe Light Water Reactor Accidents

A methodology based on chemical thermodynamics has been developed to treat the transport of volatile fission products (FPs) through the core and the primary system. The FPs considered are cesium, iodine, tellurium, strontium, and ruthenium, which may pose the major biohazard in postulated severe accidents in light water reactors. The vapor transport of FPs depends on the volatilities of the chemical compounds that are formed in the carrier gas environment in which the FPs are released and transported. Chemically stable forms were evaluated by minimizing the total free energies of the FP/ fuel/gas environment systems. Many gaseous species for each FP were considered and their partial pressures calculated over a range of temperatures (600 to 3000K), the carrier gas environments (total pressure and ratio of H/sub 2//H/sub 2/O), and the total amount of FPs in the system. It was found that the major dependence of the concentration of the FPs was on the gas temperature, and a model was developed to predict the source of volatile FPs. The model showed that the FPs leaving the core region would condense in the cooler regions of the upper plenum and/or the primary system either on the cold surfaces or be transported furthermore » as aerosols.« less

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Impact of Secondary Effects on the Reduction of Fission Product Source Terms in Class IX Reactor Accidents

This paper examines the potential radiological consequences of a hypothetical line-break, core melt type of accident (Class IX). This paper concludes that the potential for catastrophic environmental release of radioactive material in case of a major reactor accident seems to be greatly suppressed by inherent processes provided by the coolant, the primary system, and the containment, without significant reliance on engineered safety features. 18 refs.

Fission Product Source Terms for the Light Water Reactor Loss-of-Coolant Accident

Models for cesium and iodine release from light water reactor (LWR) fuel rods defected in steam were formulated based on experimental fission product release data from several types of defected LWR...

Recent Improvements in Identifying Fission Product Sources in the Experimental Breeder Reactor II

The detection of a xenon tag released from an element with breached cladding has proved to be the best method of identifying a source of fission product release in the Experimental Breeder Reactor II. From experience gained from December 1974 to September 1977 in identifying the sources of 25 major releases, allowance can now be made for the composition changes in xenon tags brought about by irradiation and for the presence of more than one tag in the cover gas. This experience has also shown that calculated ratios of fission gas to tag gas in subassemblies can be used in identifying a source of release. An ancillary method, which uses the measured /sup 131/Xe//sup 134/Xe ratio in the cover gas, can also determine whether or not the source is uranium. The success of these recent improvements is reflected in the short time to identify a source and in the small number of other sources, frequently none, that are suspect.

Nuclear generated plasmas in noble gas thermionic converters

The generation of a plasma by fission fragment ionization in noble gas thermionic converters was investigated in a series of inpile experiments at the University of Michigan reactor. The plasma generated in Ne : A (1000 : 1) at a pressure of 20 mm Hg was investigated in a plane parallel diode with electrically heated emitter and collector of barium impregnated tungsten. The fission fragment source was uranium foil mounted on a third electrode. As calculated from diode V- I characteristics, ion densities of order 10 10 cm −3 were generated by fission fragments in the space surrounding the converter, and ion densities of order 10 11 cm −3 were present in the interelectrode gap due to the formation of ion trapping sheaths at the electrode surfaces during normal converter operation. The effect on converter operation of cesium generated in the fission process was investigated in a tube with a diode structure and filling as above, but with a fission product source shielded from the interelectrode space. No cesium effects were observed during the 11 1 2 hr of tube operation. The ion density due to the reactor gamma flux was calculated from the V- I characteristic to be of the order 10 8–10 9 cm −3. Fission product poisoning of the collector was observed at temperatures less than 1000°K; the collector could be reactivated. To investigate the plasma generated by fission fragments in Xe at a pressure of 8 mm Hg, a plane parallel diode was constructed with a nuclear heated emitter of uranium carbide-rhenium and a barium impregnated tungsten collector. The emitter operating temperature was 1950°K in a neutron of 5 × 10 12 cm −2 sec −1. Discontinuities were observed in the forward current characteristic between 6–7 V. The effects between 6 and 7 V were attributed to the formation of xenon metastable atoms in the collector sheath which could then produce additional ionization. The discontinuity at 10 V was attributed to Xe breakdown. The ion density calculated from V- I characteristic before the first discontinuity was of order 10 9 cm −3.