VINSAP Tool for Visualization of Severe Accident Progress

The contribution provides information about the development of a system for visualization of NPP severe accident progress. This visualization is under development in cooperation of UJV Rez, a.s. and Czech Technical University in Prague. The project is supported by the Technology Agency of the Czech Republic and is planned to be solved from 2015 to 2017. The visualization uses results of an analytical code MELCOR for evaluation of the NPP severe accident progress. The visualization firstly reads MELCOR results, transforms them to a suitable format for quick processing and provides graphical screens with reactor components that could demonstrate the progress of the evaluated severe accident. The visualization can even provide parallel presentation of more different scenarios of the severe accident. The system is planned to be used for training of NPP staff to handle severe accidents. In the first year of the project solution (2015), the software for MELCOR data transformation, next for providing information about transformed data were developed. In the following year (2016), software for creation of graphical screens with reactor components and software for severe accident progress presentation is creating. In the final year of the project (2017), thorough testing is going to be carried out, and the applicability of the visualization for a practical use during a NPP staff training is going to be verified.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

The Study of CAP1400 Core Melt and Relocation for IVR Analysis

IVR (In-Vessel Retention) strategy is designed as the key severe accident mitigation feature for CAP1400. This paper studies the core melt and relocation progression, which is the base of the melt pool analysis and assessment in the plenum. The MAAP and CFD code are used together to obtain the main insights of the phenomena during core melting. The MAAP code is adopted to have an overall understanding of the progress with the lumped calculation, while the CFD code is used as the tool to study the local failure of the complex structure such as shroud and barrel with finite element simulation. Based on the analysis, the core will heat up after uncovered, and the upper region will melt first to form the core melt pool, as there is still water exist in the active fuel region at the time of upper part rods melting, the debris would be refrozen to form crust to block the relocation. As the melt pool increasing, the shroud is melt-through from the corner, and melts would drop to fill the gap volume between the shroud and barrel before relocation to lower plenum. Furthermore, the barrel will be melted later and the debris relocation to the lower plenum from the core sideward. The melts will touch the lower core support plate before water in the plenum depleted, which provides large mass of metal to be melted into the pool, avoiding large heat flux to challenge the RPV in the pool forming stage.

Similar but Different: A Top-Down Benchmarking Approach to Investigate Nuclear Decommissioning Projects

Project management literature has, until now, mainly focused on new build and only in the last decades the issues of decommissioning (mega) projects has arisen. To respond to this changing environment, project management will need to understand the challenges of decommissioning projects. Decommissioning projects within Oil & Gas, Chemical and Nuclear sectors are characterized by high costs, long schedules and uncertainty-based risks. The budget for Nuclear Decommissioning Projects and Programmes (NDPs) are subject to well publicized increases and, due to their relatively recent emergence, complexity and variety, key stakeholders lack a full understanding of the key factors influencing these increases. Benchmarking involves “comparing actual or planned practices [...] to identify best practices, generate ideas for improvement” [1] and offers significant potential to improve the performance of project selection, planning and delivery. However, even if benchmarking is the envisaged methodology to investigate the NDPs characteristics that impact on the NDPs performance, until now, it has only been partially used and there is a huge gap in the literature concerning benchmarking NDPs. This paper adapts a top-down benchmarking approach to highlight the NDPs characteristics that mostly impact on the NDPs performance. This is exemplified by a systematic quantitative and qualitative cross-comparison of two major “similar-but-different” NDPs: Rocky Flats (US) and Sellafield (UK). Main results concern the understanding of the alternatives of the owner and/or the contractors in relation to (1) the physical characteristics and the end state of the nuclear site, (2) the governance, funding & contracting schemes, and (3) the stakeholders’ engagement.

Preparation and Co (II) Removal of ZnO/Poly-Piperazineamide Nanofiltration Membranes

A series of nanofiltration membranes were prepared by interfacial polymerization of piperazine and terephthaloyl chloride on the surface of polyacrylonitrile (PAN) ultrafiltration membranes. ZnO nanoparticles were incorporated in the active separation layer to modify the performances of the membranes. The preparation conditions as the monomer concentration, dosage of nano-ZnO particles and the reaction time on removal of a simulated radioactive nuclide Co (II) were investigated. Fourier transform infrared in attenuated total reflection mode verified the formation of polyamide on the PAN ultrafiltration membrane. The scanning electron microscope images showed that the nano-ZnO particles can homogeneously fixed on the membrane surface. The retention of Co (II) increased with increasing the dosage of nano-ZnO in the range of 0∼0.03 g. Further adding more nano-ZnO, the rejection rate of Co (II) first decreased and then increased. The concentration of piperazine and terephthaloyl chloride showed similar effect on removal of Co (II) ion. 5 minutes polymerization time was sufficient to form an active separation layer on the substrate membrane which changed the separation mechanism from ultrafiltration to nanofiltration. The separation performance of NF3 prepared by the following conditions was optimum: 0.03g nano-ZnO, 0.6 wt% piperazine, 0.5 wt% terephthaloyl chloride, and the reaction time was 15 min. The rejection rates of 1000 mg/L Na2SO4 and Co2+ in CoCl2 solution were 90% and 75% respectively. The Co (II) removal rate can be increased to nearly 90% by using ethylenediaminetetraacetic acid disodium salt. Increasing the operation pressure or the feeding concentration of Co (II) can also improve the performances of the membranes in this experiment.

Radiation Shielding Analysis of a Spent Fuel Dry Storage Cask for FA300 Spent Fuel Assemblies

Spent fuel dry storage technology is one of the most important intermediate storage technologies for spent fuel, because of its high security, good economic and easy to expand the scale. This article aims at designing a spent fuel dry storage cask which can contain 21 FA300 spent fuel assemblies. The spent fuel dry storage cask is designed as concrete cask structure, which has the advantages of low manufacturing cost and simple manufacturing technology. Ventilation channels are designed for heating transfer, because the concrete is not a good thermal conductivity material. And labyrinth structure is designed for the ventilation channel to reduce the cavity streaming. Radiation sources in spent fuel assemblies are mainly produced from fission products, actinides and their daughters located inside the effective fuel region, and other activation products in structure materials, which are calculated by ORIGEN. The source and geometry of this problem are complex, and this is a real world deep penetration and streaming problem. Discrete ordinate method has great advantage in solving the deep penetration problem. Based on three-dimensional discrete ordinate code TORT, radiation shielding design method for spent fuel dry storage cask is studied, including main shield cask, cover lid, and ventilation channel. The results show that this spent fuel dry storage cask containing 21 FA300 spent fuel (cooling time: 10 years) assemblies can satisfy the requirement of dose rate limits in GB18871.

Development and Application of the Device for Cutting and Disintegration of Radioactive Contamination Glove Box

In the process of decommissioning of the nuclear facilities, abandoned and contaminated equipments or devices often need to be dismantled and cut into proper pieces in order to facilitate subsequent treatment and disposal. Since the nuclear facilities were places where radioactive operations were frequently executed, gloveboxes should be such typical abandoned and contaminated devices. Usually, gloveboxes were cut into proper pieces by kinds of tools which was chosen depending on the thickness of stainless steel from different parts of gloveboxes. This traditional cutting was laborious and high concentration of harmful aerosols and gases would be created during the cutting operation. In order to develop more advanced cutting and disintegration ways, a cutting and disintegration device was developed using abandoned and contaminated gloveboxes as operation objects in this work. During the design of this device, the operation convenience, operation exactitude and the protection for operators were fully considered. Also, the hot verification test was carried out. Based on the verification test results, the cutting and disintegration device was reliable and could meet the design requirements, which precisely executed various movements required in the hot verification test. Due to the application of remote operation and advanced cold cutting as the main cutting way in the development of this cutting and disintegration device, radioactive aerosols and harmful gases created during operation of this device were obviously declined compared to past cutting devices, which was of great importance to the health of operators. This work can provide technical support for the development of other similar devices applied in nuclear facility decommissioning.

Research Progress of Red Oil Explosion Accidents in Nuclear Fuel Reprocessing Plant

Since Westinghouse Savannah River Company (WSRC) of America first applied PUREX process in 1954, PUREX process is always the top priority in nuclear fuel reprocessing plant. And this process is based on liquid to liquid extraction with TBP as the extractant. TBP is irreplaceable in the development of PUREX process in nuclear fuel reprocessing, its advantages are well recognized. However TBP does have some disadvantages such as formation of red oil, which will appear in the system of high nitric acid concentration and heavy metal nitrate, once the red oil forms, it can lead a exothermic runaway decomposition in reasonable conditions, such as exceeding a certain temperature (typically 130°C) or high acid concentration. If gas products and energy released from the decomposition reaction could not be exported in time, it will lead to vessel overpressure and caused violent explosion accidents. By now, it has happened 6 times so-called red oil explosion accidents worldwide, resulting in different degrees of equipment and construction damage and environmental contamination. From 1953 to now, research related to red oil has never stopped. WSRC, Hanford Company, Oak Ridge National Laboratory and Los Alamos National Laboratory of America have conducted many studies, as well as some research institutions from Russia, UK, France and India. Defense Nuclear Facilities Safety Board of America issued a technical report in 2003, preventive measures for red oil explosion were established in this report, and these measures provided good practice experience and reference for other countries, and the temperature condition (⩽130°C)and nitric acid concentration (⩽10M)for preventing red oil explosion are employed in some countries which has built the reprocessing plant. Nevertheless, research conclusions and knowledge of red oil vary from country to country. Especially, Kumar and Smitha etc. conducted several experiments in adiabatic condition in recent years, and investigation on stability of TBP - nitric system was made, the results indicated that the red oil runway reaction will happen even in lower temperature and lower nitric acid concentration in contrast with the reported value, and they thought it would need a further study to assess the validity of present preventive measures, and to rebuild the safety limits for preventing red oil explosion in the operation of nuclear fuel reprocessing plants. In this paper, related research results of red oil explosion accidents were combed, and the characters of study work of different periods were summarized, and definition, formation conditions of red oil, pathway of runaway reaction, control and preventive measures for preventing red oil explosion of different countries were analyzed and compared, as well as the new viewpoints of recent literatures. And some research ideas for future investigation based on present work were also proposed.

HEPA Filter Clogging and Volatile Material Release Under Solvent Fire Accident in Fuel Reprocessing Facility

The new licensing standards were further improved by taking into account of lessons learned from the Fukushima-Daiichi nuclear accident, and countermeasures against severe accidents were newly required as regulatory items, where severe accidents were defined as serious accidents that occur under conditions exceeding design bases. Organic solvent fire in cell was defined as one of the severe accidents in nuclear fuel reprocessing facilities, which should be investigated, in order to establish methods for evaluating effectiveness of the countermeasures. One of the combustibles in the fire accident at reprocessing facilities is the organic solvent composed of 30% tributyl phosphate (TBP) and 70% dodecane. When the solvent burns, aerosol of soot and radioactive substances are released inside the facility. The aerosol causes a clogging of high-efficiency particulate air filters (HEPA filters) in a ventilation system of the facility, which increases a differential pressure of the filters. We have performed combustion tests simulating the fire accident. As one of interesting results of the tests, we observed, when most of dodecane in the solvent was burned out, a rapid increase in a differential pressure of a HEPA filter, which may cause its rupture. We also found a small amount of RuO4 release from the burning solvent, which can pass through HEPA filters due to its volatility. These phenomena should be adapted in the effectiveness evaluations of the countermeasures against the fire accident.

Study on Periodic Shielding Performance Measurements of Spent Fuel Transportation Package

The radioactive activity of spent nuclear fuel is high, and the transportation safety is concerned by public and specialist. The periodic radiation shielding performance measurements of spent fuels package is important content to ensure transportation safety of spent fuels. The radiation shielding performance of package must meet the requirements of “Regulations for the safe transport of radioactive material” (GB11806-2004). However, some of the problems and difficulties reflected in practice need to be solved, such as the measurements results of neutron radiation level of spent fuels package outer are not always reliable. In this paper, the periodic shielding performance measurements of one type of spent fuel transportation package are presented. The monitoring results of using both the neutron multi-sphere spectrometer and portable neutron measurement instrument are compared, and the Monte Carlo simulation is done to verify the measurements results. Some factors are discussed, and an optimized scheme is recommended.