Conditioning Operation by an Encapsulation and Shielding of Spent Sealed Radium Sources

In order to avoid accidents that could result from an improper storage of spent radium sources, it is necessary to condition and store them safely. The program for a safe conditioning of spent radium sources by the IAEA (International Atomic Energy Agency) has been established to assist the developing countries. The main object of this paper is to summarize the technology that was adapted by the IAEA for the conditioning of spent radium sources in the national inventory of Ra-226 sources in member states and the actions performed by the Korean expert team as a part of the IAEA’s project titled as ‘Radium Conditioning in Southeastern Asia’. The whole inventory of the spent radium sources of 8,671.13mCi was safely conditioned by the Korean expert team according to the guidelines under the supervision of the IAEA’s technical officer. The 1,821 sources of spent radium were encapsulated, welded, and conditioned into 18 concrete-shielded drums.

Digital Instrumentation and Control Failure Events Derivation and Analysis by Frame-Based Technique

A frame-based technique, including physical frame, logical frame, and cognitive frame, was adopted to perform digital I&C failure events derivation and analysis for generic ABWR. The physical frame was structured with a modified PCTran-ABWR plant simulation code, which was extended and enhanced on the feedwater system, recirculation system, and steam line system. The logical model is structured with MATLAB, which was incorporated into PCTran-ABWR to improve the pressure control system, feedwater control system, recirculation control system, and automated power regulation control system. As a result, the software failure of these digital control systems can be properly simulated and analyzed. The cognitive frame was simulated by the operator awareness status in the scenarios. Moreover, via an internal characteristics tuning technique, the modified PCTran-ABWR can precisely reflect the characteristics of the power-core flow. Hence, in addition to the transient plots, the analysis results can then be demonstrated on the power-core flow map. A number of postulated I&C system software failure events were derived to achieve the dynamic analyses. The basis for event derivation includes the published classification for software anomalies, the digital I&C design data for ABWR, chapter 15 accident analysis of generic SAR, and the reported NPP I&C software failure events. The case study of this research includes (1) the software CMF analysis for the major digital control systems; and (2) postulated ABWR digital I&C software failure events derivation from the actual happening of non-ABWR digital I&C software failure events, which were reported to LER of USNRC or IRS of IAEA. These events were analyzed by PCTran-ABWR. Conflicts among plant status, computer status, and human cognitive status are successfully identified. The operator might not easily recognize the abnormal condition, because the computer status seems to progress normally. However, a well trained operator can become aware of the abnormal condition with the inconsistent physical parameters; and then can take early corrective actions to avoid the system hazard. This paper also discusses the advantage of Simulation-based method, which can investigate more in-depth dynamic behavior of digital I&C system than other approaches. Some unanticipated interactions can be observed by this method.

Gamma Ray Attenuation Coefficient Measurement in Energies 1172 keV and 1332 keV for Neutron Absorbent Saturated Solutions

The compounds, Na2B4O7, H3BO3, CdCl2 and NaCl and their solutions, attenuate gamma rays in addition to neutron absorption. These compounds are widely used in shielding of neutron sources, reactor control and neutron converters. Mass attenuation coefficients of gamma related to saturated solutions of the above four compounds, in energies 1172 keV and 1332 keV have been measured by NaI detector and agree very well with the results obtained by Xcom code. Experiment and computation show that, H3BO3 has the highest gamma ray attenuation coefficient among the aforementioned compounds.

Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE–II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m3 RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa.

Results of the QUENCH-L2, DISCO-L2, and COMET-L2 Experiments Performed Within the LACOMERA Project at the Forschungszentrum Karlsruhe

The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a 4 year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities are being used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarizes the main results obtained in the following three experiments: QUENCH-L2: Boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L2: Fluid-dynamic, thermal, and chemical processes during melt ejection out of a breach in the lower head of a pressure vessel of the VVER-1000/320 type of reactor. COMET-L2: Investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase.

Study for Nuclide Transfer Ratio of Particles Generated by Thermal Cutting

In dismantling of nuclear facilities, secondary products would disperse into air and water. Thus, it is necessary to assess the effective dose for the public due to the gaseous and liquid radioactive wastes discharged. Especially for the cutting of a highly activated reactor core structures, the material may release special nuclides, some of which can be metal with low melting temperature and relatively high vapor pressure. In this case, the vaporized metal easily transfers to the particles instead of leaving in the dross and would make significant impact to the work place and environment. A pressure-tube type reactor often utilizes zirconium alloy as core structure, which may contains nuclides with high vapor pressure such as radioactive tin or antimony caused by activated tin. In this study, the radioactivity transfer ratio for thermal cutting in air has been investigated by cold tests. The tests have been carried out in an air-tight clean house and used a dust sampling system. The test piece was zirconium and niobium 2.5% alloy with some impurities, which was non-radioactive spare material for the pressure tube of Fugen NPS. Plasma cutting method was carried out with 200 A electric current. The particles were separated by one micrometer size with membrane filters and the material was investigated with an Energy Dispersion X-ray Analyzer (EDX) and the shape of the particle was observed with a Scanning Electron Microscope (SEM). The results showed that there was almost no tin in the dross because most of the tin was evaporated and transferred to aerosol. However, the ratio of zirconium, niobium and iron in the aerosol was not as high as that of tin. The vaporization pressure of antimony is higher than that of tin. The tin can be easily activated by neutron and becomes radioactive antimony. Because Zircalloy-2 or 4 contains ca. 1 percent of tin, irradiated Zircalloy contains significant radioactive antimony (Sb-125). Preparatory analyses showed that the Sb-125 to the environmental impact was comparable with other nuclides. For this reason a test for taking data of the transfer ratio of antimony by using irradiated material has been also carried out.

Experimental Evaluation of Pool Fire Suppression Performance of Sodium Leak Collection Tray in Open Air

In the event of sodium leakage from heat transfer circuits of fast breeder reactors (FBR), liquid sodium catches fire in ambient air leading to production of flame, smoke and heat. One of the passive fire protection methods involves immediate collection of the leaking sodium to a sodium hold-up vessel (SHV) covered with a sloping cover tray (SCT) having a few drain pipes and one vent pipe (as in Fig. 1). As soon as the liquid sodium falls on the sloping cover tray, gravity guides the sodium through drain pipes into the bottom tray in which self-extinction occurs due to oxygen starvation. This sodium fire protection equipment called leak collection tray (LCT) works without the intervention of an operator and external power source. A large number of LCTs are strategically arranged under the sodium circulating pipe lines in the FBR plants to serve as passive suppression devices. In order to test the efficacy of the LCT, four tests were conducted. Two tests were with LCT having three drain pipes and rest with one. In each experiment, nearly 40 kg of hot liquid sodium at 550 °C was discharged on the LCT in the open air. Continuous on-line monitoring of temperature at strategic locations (∼ 28 points) were carried out. Colour videography was employed for taking motion pictures of various time-dependent events like sodium dumping, appearance of flame and release of smoke through vent pipes. After self-extinction of sodium fire, the LCT was allowed to cool overnight in an argon atmosphere. Solid samples of sodium debris in the SCT and SHV were collected by manual core drilling machine. The samples were subjected to chemical analysis for determination of unburnt and burnt sodium. The results of the four tests revealed an interesting feature: LCT with three drain pipes showed far lower sodium collection efficiency and much higher sodium combustion than that with just one drain pipe. Thermal fluctuations in temperature sensor located near the tip of the drain pipe have indicated that transient freezing and remelting processes are responsible for this phenomenon. Moreover comparison of test results between present and earlier experiments has revealed that the LCT with funnel shaped SCT is superior to that with boat shaped SCT.

The Fate of Hazardous Materials in the Biosphere

Although nuclear power appears to be expanding as a major global energy source, the disposal of radioactive waste from the nuclear fuel cycle still poses formidable challenges to the full expansion of the nuclear enterprise. The perception that nuclear wastes represent unique and insoluble threats to humans is ill founded. The risk from these radioactive materials is comparable and many ways less severe than other more familiar hazardous materials that are ubiquitous in the biosphere. Radioactive materials decay and reduce in time unlike stable elements. Besides the reduction of radioactive materials through decay, the dilution and dispersion of all hazardous materials by natural forces and events provides the reduction required to make adequate and safe disposal of nuclear waste possible. The ultimate sink for essentially all of these hazardous wastes will prove to be the oceans with their great capacity of dilution and containment.

Evaluation and Numerical Simulation of Tsunami for Coastal Nuclear Power Plants of India

Recent tsunami generated on December 26, 2004 due to Sumatra earthquake of magnitude 9.3 resulted in inundation at the various coastal sites of India. The site selection and design of Indian nuclear power plants demand the evaluation of run up and the structural barriers for the coastal plants: Besides it is also desirable to evaluate the early warning system for tsunamigenic earthquakes. The tsunamis originate from submarine faults, underwater volcanic activities, sub-aerial landslides impinging on the sea and submarine landslides. In case of a submarine earthquake-induced tsunami the wave is generated in the fluid domain due to displacement of the seabed. There are three phases of tsunami: generation, propagation, and run-up. Reactor Safety Division (RSD) of Bhabha Atomic Research Centre (BARC), Trombay has initiated computational simulation for all the three phases of tsunami source generation, its propagation and finally run up evaluation for the protection of public life, property and various industrial infrastructures located on the coastal regions of India. These studies could be effectively utilized for design and implementation of early warning system for coastal region of the country apart from catering to the needs of Indian nuclear installations. This paper presents some results of tsunami waves based on different analytical/numerical approaches with shallow water wave theory.

First Results of the Phebus FPT3 Test

The purpose of this paper is to provide a preliminary overview of the phenomena observed during the experimental phase of the PHEBUS Fission Product Test FPT3. This experiment was the last in the series of 5 in-pile integral experiments performed by IRSN in the PHEBUS facility operated by the CEA on the site of Cadarache. Unlike the previous tests, FPT3 used boron carbide as absorber material instead of silver-indium-cadmium, so varying an important parameter impacting physico-chemical phenomena. FPT3 test course was in agreement with the pre-defined test protocol, including a 8,5-day irradiation phase, a fuel bundle degradation phase which lasted less than 5 hours and a 4-day long-term phase that consisted of an aerosol stage dedicated to the analysis of aerosol deposition mechanisms inside the containment vessel and a chemistry stage devoted to the analysis of the iodine chemistry. During the experiment, both the on-line instrumentation and the periodic samplings worked quite well. The fuel degradation progress could be analysed through both temperatures inside the bundle and gaseous concentration measurements performed in the circuit and inside the containment vessel. Some major events, like fuel clad and absorber rod failures or material relocations, were clearly correlated to both bundle and circuit instrumentation signals. The post test non destructive examinations of the fuel bundle (X-radiography, X- and γ-tomographies and γ-scanning) allowed to compare FPT2 and FPT3 bundle final degradation states. On-line γ-detector measurements coupled with numerous post test gamma-counted sequential samplings help for the characterization of the iodine behaviour inside the containment vessel during the degradation and the long term phases. The whole set of measurements appears self-consistent and provides new data for the iodine solubility inside the sump, the iodine gaseous fraction and the organic versus molecular iodine distribution inside the containment atmosphere.