Fission Product Inventory Research Articles

Estimation of radionuclide transport to reactor containment building under unprotected severe accident scenario in SFR

In the design of Sodium cooled Fast Reactor (SFR), Defense in Depth (DiD) approach is followed with multiple barriers to limit the release of radionuclides to the environment under accidental conditions. Under severe accident scenario, the multiple barriers provided as part of DiD, can fail one by one and finally the radionuclides may be released to Reactor Containment Building (RCB). Under severe accident scenario, the extent of radionuclides release into the environment has to be limited based on the permissible dose to the public at the site exclusion zone boundary. The radionuclides of concern for SFR under accidental release are noble gases and volatile fission products (Cs, I, Te). A lumped analytical methodology has been developed to estimate the source term in the core and transport of radionuclides to cover gas region/RCB under severe accident scenario in SFR. The methodology estimates the fission product inventory in the core based on saturation analysis method. In the event of severe accident, the core melts and fission products are released into the sodium pool. The transport of fission products (volatile and non-volatile) from pool to cover gas region/RCB takes place mainly by the short term phenomenon (noble gas bubble transport) and sodium spray fire release. The transport of Volatile Fission Products (VFPs) to cover gas region/RCB also takes place via the long term phenomenon (by evaporation). The evaporative transport of VFPs from hot pool to cover gas region is evaluated based on equilibrium partition coefficient (KD). The methodology is verified with the results for radionuclide inventory in SFR core, available in literature and their fractional release to the RCB. This methodology is useful for the evaluation of public dose at the site exclusion zone boundary of a typical SFR under severe accident scenario.

SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel

Predictive capabilities for simulating irradiated nuclear fuel behavior are enhanced in the current work by coupling thermochemistry, isotopic evolution and heat transfer. Thermodynamic models that are incorporated into this framework not only predict the departure from stoichiometry of UO2, but also consider dissolved fission and activation products in the fluorite oxide phase, noble metal inclusions, secondary oxides including uranates, zirconates, molybdates and the gas phase. Thermochemical computations utilize the spatial and temporal evolution of the fission and activation product inventory in the pellet, which is typically neglected in nuclear fuel performance simulations. Isotopic computations encompass the depletion, decay and transmutation of more than 2000 isotopes that are calculated at every point in space and time. These computations take into consideration neutron flux depression and the increased production of fissile plutonium near the fuel pellet periphery (i.e., the so-called “rim effect”). Thermochemical and isotopic predictions are in very good agreement with reported experimental measurements of highly irradiated UO2 fuel with an average burnup of 102GWdt(U)−1. Simulation results demonstrate that predictions are considerably enhanced when coupling thermochemical and isotopic computations in comparison to empirical correlations.Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

DFT-based prediction of fission product sorption on carbon structures under O2 ingress conditions

An isotherm based model for the prediction of Cs sorption on the carbon components of a High Temperature Reactor (HTR) under O2 ingress conditions is presented. Isotherms are derived from a thermodynamic model based on binding energies calculated using Density Functional Theory (DFT). The DFT derived isotherms are compared with isotherms obtained from experimental calculations and sources of discrepancies are discussed. A DFT only model and a second model combining DFT and experimental calculations are used to predict fission product inventories in a HTR vessel during O2 ingress conditions. Results suggest that the carbon type (i.e. graphitic vs. amorphous) plays a central role on fission product sorption and release. During normal reactor conditions (T around 1400K, low PO2) graphitic carbon will absorb a small percentage of a monolayer of Cs, while amorphous carbon will be approximately saturated at an entire monolayer of Cs. Results also indicate that, for the case of O2 ingress to the reactor’s vessel, the Cs will form Cs2O. In the case of graphitic carbon, the Cs2O will bind more weakly than Cs, leading to Cs release in the form of Cs2O during O ingress. However, the weak binding of Cs to graphite means that only small release is expected. In the case of amorphous carbon, Cs2O binds almost as strongly Cs, and so no significant change in Cs absorbed to the amorphous carbon is predicted, although the form of the absorbed Cs is predicted to be Cs2O. For the case of low release conditions, consistent with modern TRISO fuels, the core will adsorb the entire Cs inventory at normal operating temperatures. However, for high Cs release conditions, consistent with older TRISO fuels, the surface sites on the core will be saturated and most of the Cs will remain in gas form or plate out on other surfaces.

Testing of an automatic outdoor gamma ambient dose-rate surveillance system in Tokyo and its calibration using measured deposition after the Fukushima nuclear accident

An in-situ fixed point radioactivity surveillance network has been developed at the Radiation Protection Bureau, Health Canada. The network consists of a number of spectrometric NaI(Tl) detectors measuring, in real-time, ambient gamma dose-rate. The present paper describes the gamma dose-rate monitoring by one detector installed at the Canadian embassy in Tokyo during the Fukushima nuclear accident. Soil samples were collected for the measurement of fallout fission products inventories at each location where the NaI(Tl) detector was installed. The gamma-ray attenuation by the soil matrix was estimated by the information on the depth distribution of 137Cs activities. The study demonstrated that the gamma dose-rates measured by the field NaI(Tl) spectrometric method agreed well with the laboratory results estimated by the inventories of fallout fission products deposited in the soil and the vertical distribution of 137Cs in the soil.

Prediction of pellet-to-gap escape and gap-to-coolant release rates of fission products

Some nuclear power plants in Korea were found to be leaking due to debris- and grid-to-rod fretting wear (GTRF)-induced fuel failures. Evaluation models for pellet-to-gap escape rates and gap-to-coolant release rates have been developed with the use of the primary coolant activity variations. It is found that the gap-to-coolant release rates for the debris fuel failure are different from those for the GTRF fuel failure. The gap-to-coolant release rate is controlled by both the primary defect and the secondary hydride-induced defect. The secondary hydride-induced through-cracks are more effective for the release of iodine and xenon than the primary defect. The fission product inventories in the gap and the gap-to-coolant release rates can be explained by the time-dependent defect progress, the water phase around the defects, the steam in the fuel gap, the pellet-to-clad contact, and the spacer grid springs/dimples contacting the defect.

Swelling due to fission products and additives dissolved within the uranium dioxide lattice

Simulations using empirical inter-atomic potentials have been used to predict the change in volume of the uranium dioxide lattice due to the accommodation of soluble fuel additives and fission products. The incorporation of divalent, trivalent and tetravalent cations are considered. The change in accommodation mechanism for aliovalent cations between UO2 and UO2+x gives rise to markedly different defect volumes. Experimental data is in good agreement with the predictions made in this work, particularly swelling as a function of dopant concentration under different conditions. The predicted defect volumes have been combined to predict the change in lattice volume with burnup (fission product inventory) due to incorporation of these soluble species, which agrees well with swelling data from irradiated fuel.

Instant Release Fractions from Corrosion Studies with High Burnup LWR Fuel.

ABSTRACTDuring irradiation in the reactor, a fraction of the fission product inventory will have segregated either to the gap between the fuel and the cladding or to the grain boundaries in the fuel. Of these nuclides, the behavior of the fission gases is best known. The part of the inventory that is rapidly released upon contact with water is designated the instant release fraction (IRF). Previous studies have shown that IRF and fission gas release (FGR) seem to be correlated. Studies of the instant release fraction from high burnup fuel is of interest for the assessment of the safety of a geological repository.The instant release fractions of129I and137Cs from five different light water reactor (LWR) fuel rods with a burnup range of 43 to 75 MWd/kgU and a fission gas release range from 0.9 to 5.0 % were studied. Four types of fuel samples (pellet, fragment, powder and fuel rodlet) have been used in the experiments. The results show that the fuel sample preparation method has a significant impact on the release from high burnup fuel samples over the time period covered by this study. Leaching of high burnup fuel samples with fuel detached from the cladding shows the highest release. The fractional129I release from such fragment samples is similar to the FGR in the corresponding rod. On the other hand, corresponding fractional release of137Cs is lower.

High-Energy Delayed Gamma Spectroscopy for Spent Nuclear Fuel Assay

High-accuracy, direct, nondestructive measurement of fissile and fissionable isotopes in spent fuel, particularly the Pu isotopes, is a well-documented, but still unmet challenge in international safeguards. As nuclear fuel cycles propagate around the globe, the need for improved materials accountancy techniques for irradiated light-water reactor fuel will increase. This modeling study investigates the use of delayed gamma rays from fission-product nuclei to directly measure the relative concentrations of <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">235</sup> U, <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">239</sup> Pu, and <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">241</sup> Pu in spent fuel assemblies. The method is based on the unique distribution of fission-product nuclei produced from fission in each of these fissile isotopes. Fission is stimulated in the assembly with a pulse-capable source of interrogating neutrons. The measured distributions of the short-lived fission products from the unknown sample are then fit with a linear combination of the known fission-product yield curves from pure <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">235</sup> U, <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">239</sup> Pu, and <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">241</sup> Pu to determine the original proportions of these fissile isotopes. Modeling approaches for the intense gamma-ray background promulgated by the long-lived fission-product inventory and for the high-energy gamma-ray signatures emitted by short-lived fission products from induced fission are described. Benchmarking measurements are presented and compare favorably with the results of these models. Results for the simulated assay of simplified individual fuel rods ranging from fresh to 60-GWd/MTU burnup demonstrate the utility of the modeling methods for viability studies, although additional work is needed to more realistically assess the potential of High-Energy Delayed Gamma Spectroscopy (HEDGS).

Accident scenarios of the TRIGA Mark II reactor in Vienna

The safety report of the TRIGA Mark II reactor in Vienna includes three accident scenarios and their deterministic dose consequences to the environment. The destruction of the cladding of the most activated fuel element, the destruction of all fuel elements and a plane crash were considered scenarios in that report. The calculations were made in 1978 with the software program named STRISK. In this paper, the program package PC Cosyma was applied on the TRIGA Mark II reactor in Vienna and the deterministic consequences of the scenarios to the environment were updated. The fission product inventories of all fuel elements were calculated with ORIGEN2. To get meteorological data of the atmospheric condition around the release area, a weather station was installed. The release parameters were taken from the safety report or were replaced by worst case parameters. This paper focuses on two accident scenarios: the destruction of the cladding of the fuel element with the highest activity content and the case of a large plane crash. The current accident scenarios show good agreement with the calculations from 1978, hence no technical modifications in the safety report of the TRIGA reactor Vienna were necessary. Even in the very worst case scenario – complete destruction of all fuel elements in a large plane crash – the expected doses in the Atominstitut's neighborhood remain moderate.

Automated analysis for large amount gaseous fission product gamma-scanning spectra from nuclear power plant and its data mining

Based on the Linssi database and UniSampo/Shaman software, an automated analysis platform has been setup for the analysis of large amounts of gamma-spectra from the primary coolant monitoring systems of a CANDU reactor. Thus, a database inventory of gaseous and volatile fission products in the primary coolant of a CANDU reactor has been established. This database is comprised of 15,000 spectra of radioisotope analysis records. Records from the database inventory were retrieved by a specifically designed data-mining module and subjected to further analysis. Results from the analysis were subsequently used to identify the reactor coolant half-life of 135Xe and 133Xe, as well as the correlations of 135Xe and 88Kr activities.

The significance of thyroid cancer in reactor safety assessment

The risk of thyroid cancer from reactor accidents has long been a salient factor in the analysis of nuclear reactor safety. This risk arises largely from the abundance of iodine-131 in the fission product inventory. Iodine disperses widely if released into the atmosphere and could enter the human food-chain, particularly through consumption of contaminated cows' milk. Iodine entering the human body concentrates in the thyroid gland, which becomes the most highly irradiated organ or tissue. However, thyroid cancer is rarely fatal even when it is not diagnosed and treated. Although it is likely to be the greatest risk of radiation induced cancer, it may not constitute the greatest risk of death. The objective of this paper is to place this risk into its proper perspective. Published predictions and the observed consequences of reactor accidents are considered. In the Chernobyl accident, an increase in the incidence of thyroid cancer in children occurred more rapidly than expected and several hundred identified cases are more likely to have been caused by radiation from Chernobyl than from any other cause. Apart from the increased incidence of thyroid cancer, no other physical effect of radiation exposure on public health is likely to be discernible from this accident or from any other reactor accident that might occur.

Studies on the overall safety aspects during irradiation of TeO 2 in the central thimble of the TRIGA research reactor

The purpose of this study is to provide a detailed safety analysis of overall system and components in terms of their ability to provide optimum output from the irradiation of TeO 2 in the central thimble of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. It identifies safety issues relevant to 131I radioisotope production and ensures that safety analysis and design are consistent. It evaluates threats developed within the facility during the irradiation process and ultimately ensures establishment of in-core safety limits and conditions at all stages of 131I production. In-core irradiation safety not only ensures the safe operation of the reactor but also strengthens the production of radioisotopes (RI). This study attempts to review and modify all safety related events and aspects relating to RI production. The three-dimensional continuous energy Monte Carlo code MCNP is used to develop a versatile and accurate full-core model of the TRIGA core. The cross-section library and fission product inventory are generated by using NJOY and ORIGEN computer codes. The methodology to evaluate heat generation and other relevant parameters necessary to provide enough information for thermal hydraulic analysis are discussed. The neutron flux distribution inside the dry and water filled central thimble is determined in order to locate the highest neutron flux trapping position. The thermal hydraulic and safety analysis are performed by elaborate numerical analysis as well as by using GENGTC computer code. A mock-up facility has also been developed to supplement and verify the theoretically predicted results. The total energy generated during irradiation of 50 gm TeO 2 sample in dry condition is found to be 113.84 w of which 75% energy is due to neutron heating and rest of the amount is from gamma heating. Around 11.28 w of heat energy is also generated in the quartz vial. When the total generated-heat transfer is considered through conduction and radiation mechanisms, the calculated temperature of 50 g of TeO 2 reaches at 970 °C. Considering simultaneous heat transfer mechanisms, (conduction, radiation and convection) the calculated maximum temperature of the 50 g of TeO 2 powder comes down at 680 °C. It may be pointed out that very high amount of heat is generated during the irradiation of TeO 2 at 3 MW reactor power in dry condition which is nearly the melting point of TeO 2 and may be termed as unsafe mode of irradiation.

Development of a method of evaluating an inventory of fission products for a pebble bed reactor

The design of the high temperature gas-cooled reactor (HTGR) has evolved and the relevant safety requirements have been defined; accordingly, the source term to be used as the basis for licensing must also be developed. However, analysis of the source term in the HTGR has not been adequately investigated and there has not been definite improvement in this respect. Because radioactivity in normal operation must be well understood, the purpose of this study is to establish a method for activity evaluation by the code combination MCNP-ORIGEN-MONTEBURNS-MOTEX. The sophisticated method, which constructs the HTR-10 core by using the unit lattice of a hexagonal prism, is developed for core modeling. The MCNP modeling is used to simulate the generation of fission products with an increase of burnup, and ORIGEN is utilized for depletion calculation of each fission product. Continuous fuel management is divided into five discrete periods for the feeding and discharging of fuel pebbles. MONTEBURNS is used for discrete fuel management. In short, this work by aid from MOTEX traces 41 isotope nuclides, the results of which seem highly probable. In addition, the inventory of actinides at the end of each cycle is also investigated. It would be informative when the waste management of spent fuel of HTGRs would be taken into account. This article lays the foundation for future work on the analysis of the source term in HTGRs and will hopefully serve as a platform from which the safety assessment of radioactive material release during accidents can be undertaken in future.

Comparative study of actinide and fission product inventory of HEU and potential LEU fuels for MNSRs

A comparative study of fuel burnup and buildup of actinides and fission products for potential LEU fuels (UO 2 and U–9Mo) with existing HEU fuel (UAl 4–Al, 90% enriched) for a typical Miniature Neutron Source Reactor (MNSR) has been carried-out using the WIMSD4 computer program. For the complete burnup, the UAl 4–Al, UO 2 and U–9Mo based systems show a total consumption of 6.89, 6.83 and 6.88 g of 235U, respectively. Relative to 0.042 g 239Pu produced in case of UAl 4–Al HEU core, UO 2 and U–9Mo based cores have been found to yield 0.793 and 0.799 g, respectively, indicating much larger values of conversion ratios and correspondingly high values of fuel utilization factor. The end-of-cycle activity of the HEU core has been found 2284 Ci which agrees well with value found by Khattab where as for UO 2 based and U–9Mo based LEU cores show 1.8 and 4.8% increase with values 2326 and 2394 Ci, respectively.

Radioisotope contaminations from releases of the Tomsk–Seversk nuclear facility (Siberia, Russia)

Soils have been sampled in the vicinity of the Tomsk–Seversk facility (Siberia, Russia) that allows us to measure radioactive contaminations due to atmospheric and aquatic releases. Indeed soils exhibit large inventories of man-made fission products including 137Cs (ranging from 33,000 to 68,500 Bq m −2) and actinides such as plutonium (i.e. 239+240Pu from 420 to 5900 Bq m −2) or 241Am (160–1220 Bq m −2). Among all sampling sites, the bank of the Romashka channel exhibits the highest radioisotope concentrations. At this site, some short half-life gamma emitters were detected as well indicating recent aquatic discharge in the channel. In comparison, soils that underwent atmospheric depositions like peat and forest soils exhibit lower activities of actinides and 137Cs. Soil activities are too high to be related solely to global fallout and thus the source of plutonium must be discharges from the Siberian Chemical Combine (SCC) plant. This is confirmed by plutonium isotopic ratios measured by ICP-MS; the low 241Pu/ 239Pu and 240Pu/ 239Pu atomic ratios with respect to global fallout ratio or civil nuclear fuel are consistent with weapons grade signatures. Up to now, the influence of Tomsk–Seversk plutonium discharges was speculated in the Ob River and its estuary. Isotopic data from the present study show that plutonium measured in SCC probably constitutes a significant source of plutonium in the aquatic environment, together with plutonium from global fallout and other contaminated sites including Tomsk, Mayak (Russia) and Semipalatinsk (Republic of Kazakhstan). It is estimated that the proportion of plutonium from SCC source can reach 45% for 239Pu and 60% for 241Pu in the sediments.

Progress of the DUPIC Fuel Compatibility Analysis—III: Reactor Safety

The compatibility of the direct use of spent pressurized water reactor fuel in Canada deuterium uranium (CANDU) reactor (DUPIC) fuel with the existing 713-MW(electric) CANDU (CANDU-6) reactor has been analyzed for large-break loss-of-coolant-accident (LOCA) scenarios such as a 55% pump suction break, a 35% reactor inlet header break, and a 100% reactor outlet header break. The compatibility was assessed for the fuel integrity against the stored energy and the radiation environmental effect resulting from the large-break LOCA. The analysis showed that the stored energy of the DUPIC fuel was below the fuel breakup energy by 32%. The environmental effect was estimated for the personal and public doses using the radiation source term obtained from one-fourth of the fission product inventory in the fuel gap of the CANDU-6 reactor, being steadily operated at full power. The analyses have shown that both the personal and population doses are below the design limits even for a postulated dual failure such as a complete loss of containment building isolation logic.

Burnup-dependent core neutronics analysis and calculation of actinide and fission product inventories in discharged fuel of a material test research reactor

The System for Analysis of Reactor Core (SARC) has been developed for burnup analysis of nuclear reactors using whole core modeling and simulation by coupling the WIMS-D4S and CITATION. The system has the capability to use 1981, 1986 and other WIMS-D libraries recently released by International Atomic Energy Agency, based on the latest cross section data. The SARC was used for the analysis of the first equilibrium core of Pakistan Research Reactor-1 using newly released JENDL3.2 based WIMS-D library. The calculated cycle length and other core related parameters were found in good agreement with the experimental results. Changes in several core parameters such as core reactivity, flux, element-wise power densities, depletion of 235U and production of 239Pu were also analysed during the cycle. It was found that all these parameters vary linearly with burnup of the core. Moreover, the actinide and fission product inventories in the discharged fuel were also computed.

05/02345 Monte Carlo modeling of KAMINI

The purpose of this study is to provide a detailed safety analysis of overall system and components in terms of their ability to provide optimum output from the irradiation of TeO2 in the central thimble of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. It identifies safety issues relevant to 131I radioisotope production and ensures that safety analysis and design are consistent. It evaluates threats developed within the facility during the irradiation process and ultimately ensures establishment of in-core safety limits and conditions at all stages of 131I production. In-core irradiation safety not only ensures the safe operation of the reactor but also strengthens the production of radioisotopes (RI). This study attempts to review and modify all safety related events and aspects relating to RI production. The three-dimensional continuous energy Monte Carlo code MCNP is used to develop a versatile and accurate full-core model of the TRIGA core. The cross-section library and fission product inventory are generated by using NJOY and ORIGEN computer codes. The methodology to evaluate heat generation and other relevant parameters necessary to provide enough information for thermal hydraulic analysis are discussed. The neutron flux distribution inside the dry and water filled central thimble is determined in order to locate the highest neutron flux trapping position. The thermal hydraulic and safety analysis are performed by elaborate numerical analysis as well as by using GENGTC computer code. A mock-up facility has also been developed to supplement and verify the theoretically predicted results. The total energy generated during irradiation of 50 gm TeO2 sample in dry condition is found to be 113.84 w of which 75% energy is due to neutron heating and rest of the amount is from gamma heating. Around 11.28 w of heat energy is also generated in the quartz vial. When the total generated-heat transfer is considered through conduction and radiation mechanisms, the calculated temperature of 50 g of TeO2 reaches at 970 °C. Considering simultaneous heat transfer mechanisms, (conduction, radiation and convection) the calculated maximum temperature of the 50 g of TeO2 powder comes down at 680 °C. It may be pointed out that very high amount of heat is generated during the irradiation of TeO2 at 3 MW reactor power in dry condition which is nearly the melting point of TeO2 and may be termed as unsafe mode of irradiation.

Gamma-ray spectrometry combined with acceptable knowledge (GSAK): A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes

Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. This paper describes these tests and their results; while the former paper in this issue presents the methodology, equipment and techniques.