In a series of previous papers already published (including icone14-89350, 89651, icone12-49432, and 49433), the author has identified that 'long cell action' corrosion plays a pivotal role in practically all unresolved corrosion issues for all types of nuclear power plants (e.g. PWR/VVER, BWR/RBMK and CANDU). Some of these unresolved issues are IGSCC, PWSCC, AOA, FAC (erosion-corrosion) and various corrosion issues in steam generators and steam turbines. In conventional corrosion science it is well established that 'long cell action' can seriously accelerate or suppress the local cell corrosion activities. Although long cell action is another fundamental mechanism of corrosion, especially in an underground corrosion (also called 'soil corrosion') arena, potential involvement of this corrosion process has never been studied for the nuclear and fossil power plants as far as the author has been able to establish. The author believes that the omission of this basic corrosion mechanism is the root cause of practically all unresolved corrosion issues for light water reactor plants worldwide. In this paper, the author further deployed this assessment, without going into details about electrochemistry, to other key corrosion issues, e.g. steam generator and turbine corrosion issues, while briefly summarizing previous discussions for completeness, as well as introducing additional experimental and theoretical evidence of this basic corrosion mechanism. Due to the importance of this potential mechanism the author is calling for institutional review activities and further verification experiments as a joint international project.

View full abstract

This paper presents the possible combination of High Temperature Gas-cooled Reactor (HTGR) technology with the Supercritical (SC) / Ultra Supercritical (USC) steam turbine technology and investigates the prospective deployments of the SC/USC HTGR power plant. Most of the existing Fossil Power Plants (FPPs) and Nuclear Power Plants (NPPs) employ steam turbine generation sets for converting thermal energy to electricity. Energy conversion efficiency can be improved by increasing the main steam pressure. Investigations on SC Water Reactor (SCWR) reveal that the fulfillment of SCWR power plant still needs further research and development. Combination of HTGR technology and SC/USC steam turbine technology widely used in FPPs presents the reality of high efficiency SC NPP. Through investigations on State-of-the-art of SC/USC steam turbine technologies, efficiencies of thermodynamic processes of HTGR plants were analyzed while comparisons were also made between SC HTGR plant and a designed Sub-Critical (Sub-C) HTGR plant. It was shown that the net plant efficiency of a SC HTGR is equivalent to that of a SC FPP, however much higher than that of a Sub-C HTGR plant. Furthermore, a SC HTGR plant has higher economic competitiveness without emission of green-house gases and pollutants.

View full abstract

Generation IV type fast reactors (FR) are expected to be commercially deployed instead of light water reactors (LWR) around 2050. Replacement of LWR to FR needs flexibility due to uncertain factors such as FR deployment rate which affects the FR fuel (plutonium, Pu) supply amount from LWR spent fuel reprocessing and the capacity of related facilities. If the FR deployment rate is as currently planned, more Pu must be prepared by expanding LWR reprocessing. If the FR deployment rate decreases, LWR reprocessing must be reduced to avoid excess Pu. To cope with this issue we proposed the innovative system called Flexible Fuel Cycle Initiative (FFCI) that has integral reprocessing for LWR and FR spent fuels. LWR reprocessing in FFCI only carries out about 90% U recovery and residual material with Pu, U (〜5%), minor actinides (MA) and fission products (FP) goes to FR reprocessing for the planned FR deployment rate. For any decrease in the FR deployment rate temporary storage will be used. Coexistence of Pu/U with MA and FP until just before Pu/U usage in the FR provides high proliferation resistance. Preliminary evaluation revealed that FFCI can reduce the LWR reprocessing capacity and LWR spent fuel storage amount compared with current plan (reference system) if the FR deployment rate decreases. Several FR deployment scenarios and countermeasures such as FFCI were investigated.

View full abstract

Bentonite is not thermodynamically compatible with the cementitious materials that will be present in the repository. However, the evolution of the bentonite buffer, induced by high-pH plumes originating from the cement-water reactions, occurs over very long time-scales. The kinetics of bentonite dissolution is a key feature to determine if the use of cement results in significant alteration of the bentonite buffer. To address this issue, a series of simulations were carried out using a reactive solute transport model. The code used in this study fully couples geochemical reactions, flow and transport, without splitting these processes, which can result in misleading solution timescales. Mineral precipitation and dissolution reactions can be modelled using user-defined kinetics while instantaneous equilibrium assumptions can be made when the reactions of interest are fast enough compared with mass transport or the data supporting the specification of a kinetic rate is insufficient. Results of the simulations suggest that provided the initial hydraulic conductivity of the bentonite buffer is as low as designed, the dissolution kinetics of the bentonite are sufficiently slow to ensure its alteration is not significant even in the case of a combination of pessimistic assumptions on the dissolution kinetics and mass transport at the cement-bentonite interface.

View full abstract

Radioactive iodine is one of the most important nuclides to be prevented for release from nuclear facilities and many facilities have off-gas treatment systems to minimize the volatile nuclides dispersion to the environment. Silver impregnated inorganic adsorbents were known as inflammable and stable fixing materials for iodine and the authors started to develop 25 years ago a kind of inorganic adsorbent that has better capability compared with conventional ones. Aluminum oxide (Alumina) was selected as a carrier material and silver nitrate as an impregnated one. Pore diameters were optimized to avoid the influence of impurities such as humidity in the off-gas stream at lower temperatures. Experiments and improvements were alternately conducted for the new adsorbent. The tests were carried out in various conditions to confirm the performance of the developed adsorbent, which clarified its good ability to remove iodine. Silver nitrate impregnated alumina adsorbent (AgA) has about twice the capacity for iodine adsorption and higher iodine removal efficiency at relatively high humidity than conventional ones. The AgA chemically and stably fixes radioactive iodine and fits the storage and disposal of used adsorbent. AgA is now and will be applied to nuclear power plants, reprocessing plants, and research facilities.

View full abstract

The solution manifold of the system of nonlinear differential equations representing a BWR needs to be examined to understand its dynamical behavior. In particular, stable or unstable fixed points and stable or unstable oscillatory solutions (or turning points/saddle node bifurcations) are of paramount interest. In this framework integrated BWR (system) codes and simplified BWR models (reduced order models, ROM) are used together to reveal the solution manifold of the nonlinear differential equations describing the system. This work is a continuation of the previous work at the Paul Scherrer Institute (PSI, Switzerland) and University of Illinois (USA) on this field. The ROM developed at PSI was upgraded by introducing the recirculation loop, subcooled boiling and modification to the feedback reactivity calculation. The upgraded ROM has been coupled with the BIFDD code which performs semi-analytical bifurcation analysis. The methodology and some results of the semi-analytical bifurcation analysis of the modified ROM will be demonstrated and discussed. The stability boundary (SB) and the nature of the Poincare-Andronov-Hopf bifurcation (PAH-B) are determined and visualized in appropriate two-dimensional parameter maps. Furthermore, for independent confirmation of the results, numerical integrations of the ROM differential equations have been carried out in the MATLAB environment.

View full abstract

When a crack occurs in a structure, the crack may grow until the structure breaks. Therefore, accurate and simple evaluation of crack behavior is important. Fracture Mechanics Parameters are used to quantitatively evaluate crack behavior. Such accurate evaluation is attained by using precise analysis methods such as the finite element method. However, precise analysis can require enormous computational costs, and thus several simplified formulas have been developed. The present authors are currently developing an engineering tool to facilitate the application of the numerous simplified formulas that have been reported. Using this tool, anyone can evaluate fracture mechanics parameters by performing simple operations. Moreover, the effectiveness of the proposed tool is expected to increase if the system can handle more realistic crack problems. Therefore, we have implemented two important functions in this system. One is a function to perform fatigue crack propagation analysis, and the other is a function to evaluate fracture mechanics parameters under an arbitrary distribution of stresses. The present paper also demonstrates the effectiveness of the newly developed system.

View full abstract

To response large scale disaster, for example hurricane, earth-quake, or terrorist attack, it is usual case that multi-organizations, such as government (central or local), commercial entities, media, and public are involved. How to make these inter- or intra dependent organizations cooperate timely and accurately during crisis is the primary concerns. In this study, we model multi-organizations cognitive process, which introduced organizational factors, individual factors and focused on how to effectively incorporate human teams into a socio-technological system. Based on it we develop a multi-agent emergency response system. The test simulation is carried out using scenarios extracted from real emergency drills.

View full abstract

To aim at the application of a promising amidic extractant, N,N,N',N'-tetraoctyldiglycolamide (TODGA), to the partitioning process of HLW, radiolysis studies of TODGA are performed in comparison with the relevant amidic extractants. TODGA has high acid resistance toward nitric acid, whereas amide-bonds tend to be broken by irradiation with γ-rays. N,N-Dioctylamine, N,N-dioctyldiglycolamic acid, and various N,N-dioctylmonoamides are mainly formed after irradiation of TODGA. While the concentration of TODGA in n-dodecane decreases with dose, the high extractabilities of TODGA for tri- and tetra-valent actinide ions at high acidity are maintained even after irradiation. With regard to the representative fission product ions, there is no remarkable change in the extractabilities of TODGA even after irradiation.

View full abstract

Characterization of C-14 in PWR Radioactive Wastes has been researched and formation mechanisms of C-14 have been discussed. It was found from the research results that the chemical formation of C-14 existed in primary coolant was organic and was low molecule compounds which are soluble in water. On the other hand, C-14 existed in condensate waste liquid and existed on solid wastes were insoluble in water and chemically stable. The insoluble C-14 component was considered to be produced by activation reaction between neutron and substances with nitrogen. Those were thought to be decomposition substances escaped from high molecular organic materials, such as mixed bed resin or diaphragm seal etc.

View full abstract

Nuclear facilities produce highly contaminated organic radioactive waste such as ion exchange resins for water purification in a nuclear plant. It is desired that these wastes will be decomposed to reduce the volume and become stable. Toshiba has developed an ion exchange resin treatment system using supercritical water. The supercritical water whose temperature and pressure exceed 647K and 22MPa is an excellent solvent for organic compounds, which can be decomposed. Actual plant size apparatus was constructed with a treatment capacity of 1 kg ion-exchange resin per hour. The reactor consists of an outer vessel and inner vessel, and the latter one was made of titanium that has anti-corrosion properties against sulfuric acid generated from resin decomposition. The reactor has a structure whereby the pre-heating part, the reaction part with a capacity of 0.025 m^3 and the cooling part are unified. The ratio of cation and anion resin is equal to one, and the concentration of resin in the slurry was 11wt%. The slurry of 9 L/h and air of 8 Nm^3/h were supplied to the reactor, and more than 99.9% of ion-exchange resin was decomposed at 723 K and 30 MPa.

View full abstract

Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to improve countermeasure against severe accidents (SAs). At the same time, we have adopted the simplification of the overall plant systems in order to improve economy. However, since one of goals of the AB1600 is to improve economy while retaining the safety performance equivalent to the current ABWR, we studied the possibility of the further safety simplification. By giving the role of coping with DBEs to passive safety system, we could change the residual heat removal system (RHR) into non-safety grade. And then, by adopting the battery-operated pump or turbine-operated pump as the emergency core cooling system (ECCS) pump, we could change the emergency on-site AC power generator such as the diesel generator into non-safety grade. As a result, the reactor building closed cooling water system (RCW) and the reactor auxiliary sea water system (RSW) had no requirement of safety-related function.

View full abstract

On the earth, there are many environmental problems. For example, rapid increase of world population causes the enormous consumption of fossil fuel and emission of CO_2 into the global air. Now, mankaind faced to deal with these serious problems. One solution for these problems is utilization of nuclear reactors. Currently, about 65% of thermal output of a nuclear reactor is thrown away to the sea or the atmosphere through a turbine condenser. When a hot-water pipeline from a nuclear plant will be constructed, the exhaust heat from nuclear reactor will able to be utilized. Therefore, authors began to study nuclear power plant system for district heating. This reactor is based on a PWR plant. Its thermal output is 10MWth and its electrical output is 3.4MW. The nuclear plant supply electricity and heat for 2000 to 3000 houses. The plant aim to supply all the energy for the adjacent pepole's life, for example, heat, electricity and hydrogen for fuel battery car. This total-energy supply system assumed to be built in Northern area such as Hokkaido in Japan. In order to develop an optimum thermal design method for the system, heat transport experiments and thermal-hydraulic calculations were carried out. Using a metal pipe covered with foam-polyurethane thermal insulator, feed-water temperature and return-water temperature was measured to evaluate heat loss. As the result, the heat loss from the hot-water temperature was very little. The thermal-hydraulic calculation method was verified and applied to actual pipeline size calculation. The result of heat loss calculation will be 0.2℃ /5km. considering these results, the best pipe specification was obtained.

View full abstract

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive because of its capability of producing high temperature helium gas and its inherent safety characteristics. Hence, the High Temperature Engineering Test Reactor (HTTR) was successfully constructed at the Oarai Research Establishment of the Japan Atomic Energy Agency. The HTTR achieved the full power of 30MW and reactor outlet coolant temperature of about 850℃ on December 7, 2001. After several operation cycles, the HTTR achieved the reactor outlet coolant temperature of 950℃ on April 19, 2004. It is the highest coolant temperature outside reactor pressure vessel in the world. This is one of the major milestones in HTGR development of high temperature nuclear process heat application. Extensive tests are planned in the HTTR and a process heat application system will be coupled to the HTTR, where hydrogen will be produced directly from the nuclear energy. This paper gives an overview of the HTTR Project focusing on the latest results from the HTTR test and the future test plan using the HTTR.

View full abstract

We are developing a new MA/Ln separation process with N,N,N',N'-tetrakis(2-pyridylmethyl) ethylenediamine, TPEN, and its derivatives for P&T technology. TPEN has six soft-donor sites as a kind of podand type molecule and can encapsulate a metal ion. TPEN has good selectivity of Am(III) from Ln(III) and has potential to establish partitioning of MA. We elucidated extraction behavior of Am(III) and Eu(III) using several organic solvents. The maximum separation factor of Am(III) from Eu(III), SFAm/Eu, was 270 at pH 4.80 with nitrobenzene as organic solvent. In synergistic extraction system, Am(III) was separated effectively with TPEN and decanoic acid diluted with 1-octanol. The synergistic extraction system was comfortable for reduction of secondary waste from the used organic solvent and operation management, because the organic phase of the extraction system was compatible with CHON principle and has low toxicity. The SFAm/Eu and the distribution ratios of Am(III) and Eu(III) were sufficient to establish the separation process. The extraction mechanism was considered ion-pair extraction. From the viewpoint of practical application, there is a problem in the characteristics of TPEN. One of them is hydrophilicity. The hydrophobicity of TPEN is improved by introducing alkyl groups. We synthesized the hydrophobic derivative N,N,N',N'-tetrakis(2-methylpyridyl) dibutylethylenediamine, tpdben, which has two n-butyl groups. Tpdben showed the maximum SFAm/Eu was 34 at pH 5.06. A hydrophobic derivative of TPEN which has selectivity of An(III) from Ln(III) was synthesized successfully.

View full abstract

In this paper, we describe a newly developed Plant Integrated Analysis System designed for plant engineers to analyze local detailed thermal hydraulic phenomena while incorporating plant response. The system can be operated through a web-based, user-friendly man-machine interface, allowing the engineers to handle analyses and to retrieve previously analyzed results from a database for easy reference. Using this analysis system, engineers can quickly recognize thermal hydraulic phenomena that occur in particular components under given plant-operating conditions. This system currently consists of one system-scale module - describing the whole plant, four large-scale modules - one describing the downcomer, one the lower plenum, one the dome of the reactor pressure vessel (RPV), and one regarding the reactor core, and two small-scale modules - one describing the core inlet and another the separator. In this system, the data of the system-scale module is calculated by a plant dynamics code, TRAC, which has been developed to evaluate the nuclear and thermal hydraulic dynamics of each bundle in the reactor core, while the data of the large- and small-scale modules are calculated by a computational fluid dynamics (CFD) code with boundary conditions basically determined by TRAC. Using this system, we have evaluated the hydrodynamic effect when the number of Reactor Internal Pumps (RIP) and Main Steam (MS) lines are reduced for the development of a medium-capacity Advanced Boiling Water Reactor (ABWR). We have also used this system to study the difference of the radial gradient of the water surface in the Reactor Pressure Vessel (RPV) between a conventional Boiling Water Reactor (BWR5) and an ABWR. Thus, it was demonstrated that the system is useful for and can support plant engineers who are unfamiliar with CFD analysis, in order to enhance plant performance.

View full abstract

This paper describes the review structural analysis for the Turbine Building of Olkiluoto 3 NPP (UMA) under finished condition when building is covered with insulated cladding and inside temperatures are maintained by HVAC system.. The main aim of the review structural analysis is to ensure the sufficient strength and deformation characteristics of the turbine building structures. The second part of this paper describes the verification structural analysis for the Cooling Water Pump Building of Olkiluoto 3 NPP (UQA). The current verification analysis assesses the validity of the Cooling Water Pump building design for serviceability limit states concentrating on crack width calculations. The introduction of construction time shrinkage joints in the turbine building deletes to the large extent the uneven shrinkage contraction in the concrete substructure. It also enables the uniform temperature expansion of the turbine building structures. The required rebar areas in the cooling water pump building were determined for long term serviceability limit states and for allowable crack widths of 0.13 mm as well as for short term serviceability limit states and for allowable crack widths of 0.26 mm. From the obtained results it can be concluded that more than 95 % from the all shell elements in the model of the cooling water pump building fulfill crack width requirements in both long term and short term load conditions with the system reinforcement of two layers of 20 mm rebars with 100 mm pitch in both directions

View full abstract

The chemical control of impurity such as hydrogen and oxygen in coolants is one of the critical issues for the development of liquid metal cooled fast reactors and self-cooled liquid breeder blankets for fusion reactors. Especially, hydrogen (isotopes) level is the key parameter for corrosion and mechanical properties of the in-reactor components. For fission reactors, the monitor of hydrogen level in the melt is important for safety operation. In addition, the control of tritium is essential for the tritium breeding performance of the fusion reactors. Therefore, on-line hydrogen sensing is a key technology for these systems. In the present study, conceptual design for the on-line hydrogen sensor to be used commonly in liquid sodium (Na), lead (Pb), lead-bismuth (Pb-Bi), lithium (Li), lead-lithium (Pb-17Li) and molten salt LiF-BeF_2 (Flibe) was performed. The cell of hydrogen sensor is made of a solid electrolyte. The solid electrolyte proposed in this study is the CaZrO_3-based ceramics, which is well-known as proton conducting ceramics. In this concept, the cell is immersed into the melt which is containing the hydrogen at the partial pressure of P_<H1>. Then, the cell is filled with Ar-H_2 mixture gas at regulated hydrogen partial pressure of P_<H2>. The electromotive force (EMF) is obtained by the proton conduction in the electro chemical system expressed as P_<H1> (melt) | solid electrolyte | P_<H2> (reference gas). The Nernst equation is used for the evaluation of the hydrogen partial pressure from the obtained EMF. The evaluations of expected performance of the sensor in Na, Pb, Pb-Bi, Pb-17Li, Li and Flibe and experimental validation at hydrogen pressures equivalent to those for the melts in the reactor conditions were carried out.

View full abstract

Data on glovebox dismantling activities in the Glovebox Dismantling Facility were analyzed to identify the work structure and the time consumed for each activity. As a result, we were able to estimate time spent for each dismantling activity. We eventually divided activities into three categories: 1) "predictable activities" with a time variable of 30% or less, 2) "suppressible activities" with little time spent compared to the total time, and 3) "unpredictable activities." In terms of these definitions, the time interval for each unit activity was evaluated. We found that almost all of the work time can be predicted within a 30% uncertainty.

View full abstract

In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data.

View full abstract

The paper presents an analysis of the effects of grain orientations on a short, kinked surface crack in a 316L stainless steel. The kinking of the crack is assumed to take place at the boundary between two neighbouring grains. The analysis is based on a plane-strain finite element crystal plasticity model. The model consists of 212 randomly shaped, sized and oriented grains, loaded monotonically in uniaxial tension to a maximum load of 0.9R_<p0.2> (240MPa). The influence that a random grain structure imposes on a Stage I crack is assessed by calculating the crack tip opening (CTOD) displacements for bicrystal as well as for polycrystal models, considering different crystallographic orientations. Since a Stage I crack is assumed, the crack is always placed in a slip plane. Results from a bicrystal case show that the maximal CTODs are directly related to the stiffness of the grain containing the crack extension. Anisotropic elasticity and crystal plasticity both contribute to this grain stiffness, resulting in maximal CTOD when the response is soft. Anisotropic elasticity can additionally increase the softness of a grain at certain crystallographic orientations. Minimal anisotropic elasticity at the crystallographic orientations with the highest Schmid factors causes the CTOD to be maximized. Presuming that the crack will preferably follow the slip plane where the crack tip opening displacement is highest, we show that the crystallographic orientation can affect the CTOD values by a factor of up to 7.7. For a given grain orientation the maximum CTOD is attained when the crack extension deflection into a second grain is between - 75.141° and 34°.

View full abstract

In boiling water reactors (BWRs), traversing incore probes (TIPs) are used for the continuous measurement of the axial distribution of neutron flux in a reactor and the calibration of local power range monitors (LPRMs). One of candidates for a substitute for TIP is gamma thermometers (GTs). Some researches have been conducted so far for the verification of applying GT to BWR plants. In the research initiated in 1996), the applicability of hardware of GT was confirmed. In the next research initiated in 2000, the applicability of GT to Core Monitoring System (CMS) in BWR plants was confirmed. Considering the results of these two researches, we manufactured GT assembly with improved structures and installed 2 of them in Kashiwazaki-Kariwa 6 (ABWR) to conduct a verification test and acquired GT signal data for 1 cycle of the plant operation. We compared the reactor power distribution calculated from TIP data with that calculated from the acquired GT signal data. The average Root Mean Square (RMS) value between these two reactor power distributions was 3.5%, and it is well consistent compared with the result of the past two researches. So we had verified the applicability of the improved GT assembly from the result of the in-plant test.

View full abstract

The distribution of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effects of the slag type, slag concentration and basicity in a plasma arc melting process. A direct current plasma arc furnace was used to melt contaminated stainless steel and real wastes from the uranium conversion plant in KAERI. The slag former used to remove the contaminants mainly consists of silica (SiO_2), calcium oxide (CaO) and aluminum oxide (Al_2O_3). Calcium fluoride (CaF_2), nickel oxide (NiO), and ferric oxide (Fe_2O_3) were added to provide an increase in the slag fluidity and oxidative potential respectively. The cerium was used as the surrogate of the uranium because the thermochemical and physical properties of the cerium were greatly similar to those of the uranium. The cerium was removed from ingot phase to slag phase by up to 99%. The removal ratio of the cerium was increased with an increase of the amount of the slag former. And maximum removal of the cerium occurred when the slag basicity was 0.82. The natural uranium (UO_2) was partitioned from the ingot phase to the slag phase by up to 95%. The distribution of the natural uranium was considerably dependent on the basicity of the slag former and the composition of slag former. The optimum condition for the removal of the uranium was about 1.5 in basicity and 15wt% of slag former. According to the increase of the amount of the slag former, the distribution of uranium oxide linearly increased due to the increase of the capacity to capture uranium oxide within the slag. Through experiments with the various slag former, we verified that the slag formers containing calcium fluoride (CaF_2) and high silica were more effective for the melting decontamination of stainless steel wastes contaminated with uranium. In the melting tests with stainless steel wastes from the uranium conversion plant in KAERI, we found that the results of the uranium decontamination were mostly similar to those of the uranium oxide.

View full abstract

Nuclear development in Japan has been continued for longer than half a century. Some nuclear facilities and power plants were built at the dawn and some of them finished their own missions. These facilities and plants are planned to be decommissioned in sequence. The Japan Atomic Power Company's (JAPC) Tokai Power Station (TPS) is the first commercial nuclear power plant in Japan, and JAPC finished its operation in 1998. Now TPS has kept safety storage period. Kawasaki Plant Systems (K-Plant) took part in the construction of TPS and had maintained some facilities until the end of the operation. Therefore JAPC and K-Plant have continued decommissioning investigations for more than 20 years. This report shows the decommissioning technology for primary loop heat exchangers (Steam Raising Unit, SRU), which K-Plant made. JAPC will apply remote cutting system and jack down system as the SRU dismantling method. K-Plant plans demolition works along JAPC's direction.

View full abstract

The blast loads have in most cases not been assumed as design basis loads of nuclear power plant buildings and structures. Recent developments however stimulated a number of analyses quantifying the potential effect of such loads. An effort was therefore made by the authors to revisit simple and robust structural analysis methods and to propose their use in the vulnerability assessment of blast-loaded structures. The leading idea is to break the structure into a set of typical structural elements, for which the response is estimated by the use of slightly modified handbook formulas. The proposed method includes provisions to predict the inelastic response and failure. Simplicity and versatility of the method facilitate its use in structural reliability calculations. The most important aspects of the proposed method are presented along with illustrative sample applications demonstrating: ・ results comparable to full scale dynamic simulations using explicit finite element codes and ・ the performance of the method in screening the existing structures and providing the structural reliability information for the vulnerability analysis.

View full abstract

Thermal aging embrittlement of LWR components made of stainless cast (e.g. CF-8 and CF-8M) is a potential degradation issue, and careful attention has been paid on it. Although welds of austenitic stainless steels have γ-δ duplex microstructure, which is similar to that of the stainless cast, examination on thermal aging characteristics of the SS welds is very limited. In order to evaluate thermal aging behavior of weld metal of austenitic stainless steel, the 316L SS weld metal has been prepared and changes in mechanical properties and in etching properties at isothermal aging at 335℃ have been investigated. The microhardness of the ferrite phase has increased with aging, while the hardness of austenite phase has stayed same. It has been suggested that spinodal decomposition has occurred in δ-ferrite by the 335℃ aging. The etching rates of δ-ferrite at immersion test in 5wt% hydrochloric acid solution have been also investigated using an AFM technique. The etching rate of ferrite phase has decreased consistently with the increase in microhardness of ferrite phase. It has been thought that this characteristics is also caused by spinodal decomposition of ferrite into chromium-rich (α') and iron-rich (α)

View full abstract

A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replaced fresh fuels. About 40 % of natural or depleted uranium undergoes fission. But presently, there are no data for material integrity under a condition of 40% burnup. In this paper, this we try to solve this proglem by the following method. Fuels burning in progress are removed from core, and recladded. During this process FP gas is removed from these fuels. Then, they are charged in the previous position of core again. Finally, 40% burnup is attained with maintaining material integrity. We performed the burn-up calculation including the above process and investigated reactor physics properties. As a result, attained maximum neutron fluence became small because of recladding. Effective neutron multiplication factor when reactor start up was 1.0015 and one just before reactor stopped was 1.003. The effects of recladding appeared small. Shape of each distribution was almost same and burnup was about 41%.

View full abstract

The 4S is a sodium-cooled small fast reactor designed to supply energy and hot water on isolated islands or in remote locations. Small reactors are often said to have disadvantages in terms of economies of scale. This problem can be overcome by making the structure simple, doing away with the need for maintenance, and designing a core that requires no refueling during reactor lifetime. Taking into account the nature of current demand from the economy and the market, there are two options for the electrical output of the 4S reactor: 10MWe and 50MWe. The core of the 10MWe 4S reactor (4S-10M) has a 30-year lifetime without refueling. Metallic fuel is employed. Burn-up reactivity loss of the 4S is regulated by neutron reflectors which surround the core. All temperature reactivity coefficients including void reactivity are kept negative during core life-time. The 4S-10M is a tall pool-type reactor. It has an intermediate heat exchanger (IHX) in the annulus space inside the reactor vessel (R/V). There are two primary electro-magnetic pumps (EM pumps) in serial under the IHX, and an air flow path on the surface of the guard vessel (G/V) as a decay heat removal system (RVACS; Reactor vessel auxiliary cooling system). The secondary sodium heated in the shell part of the IHX flows to the steam generator (SG) via the tubes of the air cooler of the intermediate reactor auxiliary cooling system (IRACS). The decay heat removal systems of 4S-10M consist of a RVACS and IRACS. Both are passive systems. To clarify the safety margin of the 4S-10M, it is important to confirm the characteristics of the two decay heat removal systems during the plant transient. The multi-dimensional thermal-hydraulic effect in the R/V could be important under such natural circulating conditions of this sort. CERES is a multi-dimensional plant dynamics simulation code for LMRs (Liquid Metal Reactors) developed by the CRIEPI (Central Research Institute of Electric Power Industry). The CERES code can handle almost all the components needed for evaluation of LMR safety. Sodium, water, lead-bismuth eutectic and lead can be handled as the coolants. The plenum in the R/V is modeled by the R-Z 2 dimension in the CERES code. The accidents of a protected loss of flow (PLOF) caused by a total station black out (TBO) and a loss of flow without scram (ULOF) were selected as the typical accidents that provide an accurate indication of the characteristics of the cooling system. The calculation clarified the thermal-hydraulic characteristics of the 4S-10M. The safety margins of the 4S were evaluated by the results of the calculations satisfied the criteria of the temperature and CDF value criteria.

View full abstract

In order to decrease the amount of aqueous liquid waste discharged from nuclear fuel reprocessing, the conversion of uranium dioxide into its nitrate using liquefied nitrogen dioxide was studied. Uranium dioxide powder was immersed in liquefied nitrogen dioxide at 313 K after a pretreatment by the oxidation of the uranium dioxide with nitrogen dioxide and air at 523 K. Seventy-nine % of the uranium dioxide, whose initial feed amount was 0.3 g, was converted into a water soluble compound. Based on an XRD analysis, uranyl nitrate trihydrate (UO_2(NO_3)_2・3H_2O) was confirmed as the product.

View full abstract

Removal of Pd, Ru and RuO_2 from molten glass was studied using liquid Cu as a collecting metal. To increase the collision frequency between Cu and these platinum group metals (PGMs), copper-ruby glass containing Cu nanoparticles was used for the removal. The glass was prepared by the reduction of a glass containing CuO with Si. Existance of the Cu nanoparticles was conformed by a measurement of the absorption peak of surface plasmon resonance at 595 nm. Another glass containing Cu paticles of around 30 μm was prepared as a control specimen. During the heating of the two glasses with PGMs, separable metal buttons were formed in both glasses. Metallic Pd and Ru were collected more than 90% in either metal buttons. There was no significant difference between the two glasses for the removal of metals. On the other hand, RuO_2 was reduced to metallic Ru and collected 83% for the ruby glass, while not more than 52% for the control glass. The use of copper-ruby glass was effective for the removal of oxides. From the leaching test of Cu into nitric acid, we also found that dispersing smaller Cu particles in the glass had an effect to make the leaching rate decrease.

View full abstract

Corrosive conditions in primary cooling systems of boiling water reactors (BWRs) are determined by oxidants, e.g., O_2, H_2O_2 and other corrosive radiolytic species. In this study, to determine the effect of H_2O_2 on the corrosion of the stainless steel in high temperature water, the electrochemical corrosion potential (ECP) and frequency dependent complex impedance (FDCI) of stainless steel specimens exposed to high temperature water containing H_2O_2 were measured. And the electric resistances of the oxide film in high temperature water were also measured. Obtained results are summarized in follows. 1) The ECP kept at same level at the wide range from 10ppb to 100 ppb of H_2O_2. 2) The sum of the anodic resistance and electric resistance of oxide film depended upon [H_2O_2], which was determined by the equivalent circuit analyses of measured FDCI. 3) The specific electric resistance of oxide film was about 3 MΩ cm, which was much smaller than the anodic resistance. It was confirmed that the dependency of FDCI on [H_2O_2] was mainly determined by the anodic resistance. 4) From the obtained results, the sensor complex for ECP and FDCI measurements was proposed to determine [H_2O_2] directly in high temperature water.

View full abstract

Equilibrium evaporation experiments have been conducted to investigate fundamental liquid-to-gas transfer behavior of volatile radionuclides polonium-210 (^<210>Po), cesium, and tellurium in a lead-bismuth eutectic (LBE) that is considered one of the promising coolant materials of Fast Breeder Reactors (FBR) and Accelerator Driven transmutation System (ADS). The experiment utilizes a conventional vapor pressure measurement method: the 'transpiration' method, which has already contributed in a sodium(Na)-cooled FBR study to the understanding of evaporation behavior for volatile fission products (cesium, iodine and tellurium) in a Na pool. Since both LBE pool test and Na pool test focus on the evaporation of the same nuclides in liquid pool, it is possible to compare the nuclides' volatility between in LBE and in Na. This paper describes first the reviewed evaporation characteristics of fission products in Na, next the evaporation test results of fission or activation products in LBE as a continuation of previous papers ICONE12-49111 and ICONE14-89187. The importance of investigating ^<210>Po evaporation is quantitatively demonstrated through specific estimation of the vapor in a cover gas region of a typical LBE-cooled system. Furthermore, comparison under a certain condition is made for the volatility of cesium and tellurium in two kinds of liquid metal coolant Na and LBE using the derived gas-liquid partition coefficients in both tests. The accumulated experimental data can serve as significant database used in accident analysis tools for safety assessment of liquid-metal-cooled nuclear reactor systems.

View full abstract

Integral Test Facility (ITF) experimental programs are relevant for validating the Best Estimate (BE) Thermal Hydraulic codes (TH) used for transient and accident analyses, design of Accident Management (AM) procedures, licensing of Nuclear Power Plants (NPP), etc. The validation process is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur for transient and/or accidents. University of Pisa (UNIPI) was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility, operated at Electrogorsk Research and Engineering Center (EREC), in the framework of the TACIS Contract 3.03.03 Part A. This paper describes the methodology adopted at UNIPI, starting form the scenarios foreseen in the final Test Matrix (TM) until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference NPP, b) the code assessment process involving the identification of phenomena challenging the code models, c) the features of the concerned ITF (scaling limitations, control logics, data acquisition system, instrumentation, etc.). An overview of all the activities performed in this respect is provided focusing the discussion on the relevance of the heat losses. This issue is particularly relevant for addressing the scaling approach related to the power and volume of the facility.

View full abstract

A series of experimental studies of the boiling liquid flow from holes of 8-60 mm diameter were conducted. The initial water pressure in the tank was 7-11.5 MPa, water temperature was 160-235℃. Various stages of jet formation were determined as well as the jet out flowing velocity, and velocity of the flow. An quasi-stationary model of boiling liquid flow was suggested. The estimated data on the density distribution of the vapor-liquid mixture in the jet were given.

View full abstract

Among nonnuclear-weapon state, Japan is the only country which is advancing nuclear fuel cycle program such as uranium enrichment, reprocessing of nuclear fuel and developing fast-breeder reactor. For gaining international acceptance, Japan has taken measures such as accepting IAEA safeguards and compliance with London guideline to untangle concerns about nuclear proliferation. However, there has been a delay in development of regal framework about control of information relating to sensitive nuclear technology. Laws and regulations which control the outflow of nuclear-related technology and service to other countries have already been established pursuant to London guideline. Meanwhile additional penalty against person who made unauthorized disclosure of sensitive information, which has potential for development of nuclear weapons, such as uranium enrichment and obligation of confidentiality has not been established by law. Official security clearance system which qualifies person to access sensitive information has not also been established but been left to nuclear companies. It is necessary to establish regal framework about control of information relating to sensitive nuclear technology as soon as possible in order not to contribute proliferation of nuclear weapons as a result by unauthorized disclosure of sensitive information by imprudent person.

View full abstract

As part of Westinghouse flawless fuel initiatives, Westinghouse has initiated companywide efforts to enhance overall quality of UO_2 pellets by enhancing pellet surface quality, pressing processes, examining pellet handling during various stages of pellet fabrication process and enhancing inspection process with primary focus of reducing pellets with missing surfaces. Benchmarking/best practices were examined among all Westinghouse and some of the Westinghouse licensee sites to minimize pellet chipping, enhance overall pellet quality and improve inspection processes. Numbers of 3 - Dimensional (3-D) finite element models were developed to analyze the impact of cladding and pellet defects on the cladding temperature, stress and strain. The Finite Element Analysis (FEA) results demonstrated that: 1. Large Missing Pellet Surface (MPS) influences the fuel rod performance characteristics such as temperature, stress and strain distributions and 2. MPS dimensions should be controlled to improve the fuel rod performance characteristic impact. Based on the FEA conclusion, review of the in-reactor fuel performance of pellets with chamfers, characterization of pellets with side chips and correlating all of the above, a new set of criteria for side chips was identified and implemented. By analyzing each process step in pellet manufacturing and performing studies to determine forces needed to induce chipping in pellets, a knowledge based handling system can be established. Critical steps were identified and analyzed for handling of pellets between the processes steps of pellet pressing and storage. Extensive process evaluations and tests have been completed in manufacturing to show that manual inspection is effective. As deemed necessary, manual inspection process was modified to further enhance current manual inspection process. Westinghouse has also initiated a development effort to replace manual inspection with an automatic inspection system for pellets. Different sites use different conversion and pelleting processes and therefore site specific improvements are being made where it would have most impact.

View full abstract

This paper is prepared in honor and in memory of the late Professor Emeritus Yasuhide Asada to recognize his contributions to ASME Nuclear Codes and Standards initiatives, particularly those related to risk-informed technology and System Based Code developments. For nearly two decades, numerous risk-informed initiatives have been completed or are under development within the ASME Nuclear Codes & Standards organization. In order to properly manage the numerous initiatives currently underway or planned for the future, the ASME Board on Nuclear Codes & Standards (BNCS) has an established Risk Management Strategic Plan (Plan) that is maintained and updated by the ASME BNCS Risk Management Task Group. This paper presents the latest approved version of the plan beginning with a background of applications completed to date, including the recent probabilistic risk assessment (PRA) standards developments for nuclear power plant applications. The paper discusses planned applications within ASME Nuclear Codes & Standards that will require expansion of the ASME PRA Standard to support new advanced light water reactor and next generation reactor developments, such as for high temperature gas-cooled reactors. Emerging regulatory developments related to risk-informed, performance- based approaches are summarized. A long-term vision for the potential development and evolution to a nuclear systems code that adopts a risk-informed approach across a facility life-cycle (design, construction, operation, maintenance, and closure) is also summarized. Finally, near term and long term actions are defined across the ASME Nuclear Codes & Standards organizations related to risk management, including related U.S. regulatory activities.

View full abstract

The some results of experimental modeling of a condensation shock phenomena in a tube are explained at the stratified flow vapor and liquid. The classification of situations is executed and the most representative kinds of flows and reasons calling condensation water hammer are determined. The method of estimation of hazard and estimation of maximum amplitude of condensation water hammer are designed at stratified vapor - liquid flow.

View full abstract

In February 2006 President Bush announced the Advanced Energy Initiative, which included the Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP). GNEP has seven broad goals, one of the major elements being to develop and deploy advanced nuclear fuel recycling technology. DOE is contemplating accelerating the deployment of these technologies to achieve the construction of a commercial scale application of these technologies. DOE now defines this approach as "…two simultaneous tracks: (1) deployment of commercial scale facilities for which advanced technologies are available now or in the near future, and (2) further research and development of transmutation fuels technologies." GE believes an integrated technical solution, using existing reactor and fuel reprocessing technologies, is achievable in the near term to accelerate the commercial demonstration of GNEP infrastructure. The concept involves a single, integrated, commercial scale, recycling facility consisting of the Consolidated Fuel Treatment Center (CFTC), capable of processing LWR and fast reactor Spent Nuclear Fuel (SNF) and fabricating Advanced Recycling Reactor (ARR) actinide fuel. The integrated facility would include a fast reactor that uses actinide-bearing fuel to produce electricity. For optimal performance, GE believes this integrated facility should be co-located to eliminate transportation between the CFTC and ARR, and enhance proliferation resistance. This Advanced Recycling Center takes advantage of previous investments by government and industry in fast reactor technology research and development. To allow for commercial acceptance, a prototypical demonstration reactor and associated fuel cycle facility will be constructed, tested, and licensed. Taking advantage of GE's NRC-reviewed modular sodium-cooled PRISM reactor, only a single reactor will be needed and the cost and risk minimized in the initial phase of the program. This paper outlines a process and a schedule to deploy the necessary domestic infrastructure to achieve the goals of GNEP. It is recognized that there are many types of public private partnerships that could demonstrate the technologies to close the fuel cycle. It is important to acknowledge the realities in the marketplace when developing an approach to advance nuclear technologies that can gain widespread commercial acceptance.

View full abstract

The paper presents two types of a passive safety containment for a near future BWR. They are tentatively named Mark S and Mark X containment in the paper. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane. The both containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is evolutionally improved. Nothing is new in the hardware but everything is new in the performance.

View full abstract

The Passive Containment Cooling System (PCCS), which remove heat released inside the containment vessel by steam condensation without using active components such as a pump during a postulated design-basis accident, have been investigated extensively. The PCCS is one of the promising safety systems which operates under the design basis accidents and protects the containment vessel by preventing heating and pressurizing. A new type of PCCS with vertical heat transfer tubes has been proposed by us, which is a closed circuit cooling system installed vertically through the containment vessel wall and cools gases inside the containment vessel and is cooled by natural convection outside of it. The objective of this study is to evaluate the mean condensation heat transfer coefficients on the external surface of the perpendicular tube for the new concept of PCCS under a non-condensable gas presence. In this paper, the effects of non-condensable gas concentration and pressure on the heat removal characteristics of the PCCS with vertical heat transfer tubes for a large dry concrete containment to mitigate containment pressure under Loss of Coolant Accident and the mean condensation heat transfer coefficients under a non-condensable gas presence are discussed.

View full abstract

In this paper, we discuss results obtained in a scoping series of experiments within a new experimental program at the Division of Nuclear Power Safety (NPS) Royal Institute of Technology (KTH). The experimental program called DEFOR was initiated to study the processes which govern debris bed formation during severe accident with core melt down and reactor pressure vessel failure at LWR plant. The objective of the present exploratory phase (DEFOR-E) is to test operational concepts, and initiate the analysis of DEFOR related phenomena. Binary oxides mixtures at different overheating were used as corium melt simulants. Sensitivity of debris bed properties to water pool depth and subcooling is discussed in the paper. The insights gained from the scoping experiments are found useful to guide the scaling rationale and design of the next series of "Snap-Shot" experiments in DEFOR.

View full abstract

In the field of digital I&C AREVA NP is focused on concepts that on the one hand make allowance for development cycles getting shorter in the technology competition and on the other hand assure a long-term system support with the ability to deliver spare parts in the long run. The system platform TELEPERM XS, which was developed especially for safety I&C application of nuclear power plants, meets requirements effectively and thus provides a great benefit for the customer. The typical applications of TELEPERM XS are in the field of Reactor Protection and ESFAS functions (Engineered Safety Features Actuation System). High demands are defined for system reliability and availability, as well as for failure prevention and tolerance. The requirements of corresponding international codes and standards of nuclear installations are also implemented in the development and engineering processes of TELEPERM XS. The system platform is integrated into a sustainable program for service life management of electronic systems and equipment. Its ongoing future-oriented development ensures the long-term availability of hardware and software components for installed TELEPERM XS applications already installed in the plants. The further development of platform and components continues to be based on the robust, service-proven TELEPERM XS architecture, with the aim of minimizing the risks associated with equipment qualification and project licensing. A further development feature is the completion and extension of TELEPERM XS applications. This continuous innovation process, combined with maximized compatibility, makes TELEPERM XS unique, and provides the basis for a sustainable system with a service life guaranteed for the long term. Within the past 10 years, the majority of all comprehensive modernization projects worldwide were implemented or are contracted using TELEPERM XS. TELEPERM XS has been implemented in two new nuclear power plants and there are orders for four more. This makes AREVA NP to the worldwide market leader in the field of safety I&C for Nuclear Power Plants. The statistical data from more than 830 million module operating hours with TELEPERM XS show extremely low module failure rates. Failures of systems implemented worldwide were not observed yet. Single failed components have been analyzed from the very beginning. Based on the reached high reliability and availability of TELEPERM XS I&C systems and the implemented self-monitoring features, utilities are able to achieve extremely low efforts for system maintenance and periodic surveillance tests.

View full abstract

Existing nuclear power plants have expended significant resources due to re-qualification, design-basis reviews, testing, unexpected failures, and high maintenance of valves. This paper presents advanced, validated predictive models and laboratory testing that can be used to assist in functional qualification. Also presented are considerations for testing and valve design beyond those found in current standards that should be incorporated into procurement specifications for valve assemblies in new generation nuclear power plants. The application of this information can significantly reduce the cost of initial functional qualification, the on-going resource requirements for valve testing and maintenance, and reduce the potential for unexpected failures and qualification concerns.

View full abstract

The mini-channel heat exchanger with micrometer or millimeter scale has the advantages on the high efficiency and high-scalability. The ratio of the heat transfer surface over the volume is larger than that of normal one. Especially, the fin inside the channel may enhance the heat transfer. In this study, to develop the high-efficiency heat exchanger, the boiling phenomena inside the fin-type mini-channel heat exchanger has been investigated. The boiling inside the mini-channel has a strong effect on the channel wall. In the previous studies, the two-phase flow inside the mini-tube or mini-channel has been analyzed with thermography and video cameras. In this study, the boiling characteristics inside the millimeter scale channel has been visualized by the high-speed camera to evaluate the efficiency of the mini-channel.

View full abstract

This paper describe about measurement for the complex and transient cross-flow over a circular cylinder with the dynamic (time resolved) PIV (particle image velocimetry) techniques. The experiment was carried out water flow tunnel with a working section of 50x50 mm, at the Reynords number 6.7 x 10^3 to 2.7 x 10^4. This circular cylinder constructed with MEXFLON resin, the end of circular cylinder is rigidly supported and the other is free. The MEXFLON is fluorine resin; its refractive index is almost same as the water with high transparency. Very high speed water flow among the test section had been clearly visualized and captured by high speed camera. The fluctuations of the flow structure also are clearly obtained with high spatial and high temporal resolution, 512x512pixel with 10,000fps. It corresponds to set up number of thousands LDV array at the test section. Consequently, we found there are asynchronous vibration between parallel-ward and perpendicular-ward to main flow.

View full abstract

Optimization of loading patterns is a very important task from economical point of view in a nuclear power plant. The optimization algorithms used for this purpose can be categorized basically into two categories: deterministic ones and stochastic ones. In the Paks nuclear power plant a deterministic optimization procedure is used to optimize the loading pattern at BOC, so that the core would have maximal reactivity reserve. To the group of stochastic optimization procedures belong mainly simulated annealing (SA) procedures and genetic algorithms (GA). There are new procedures as well, which try to combine the advantages of SAs and GAs. One of them is called population mutation annealing algorithm (PMA). In the Paks NPP we would like to introduce fuel assemblies including burnable poison (Gd) in the near future. In order to be able to find the optimal loading pattern (or near-optimal loading patterns) in that case, we have to optimize our core not only for objective functions defined at BOC, but at EOC as well. For this purpose I used stochastic algorithms (SA and PMA) to investigate loading pattern optimization results for different objective functions at BOC. In this paper I would like to present these results.

View full abstract

Recent industrial lasers and newly-developed high performance lasers will open new possibilities of laser decommissioning that we can clean-up quickly every facility of the nuclear power reactors, decontaminate each piece of the disassembled devices and sharply cut without producing a large amount of radioisotope (RI)-contaminated dusts during the decommissioning operation[1]. In the conference, we plan to report recent results of laser decommissioning trials and basic performances for cold materials using several laser types, and to discuss finally about feasibility of the laser decommissioning for Fugen nuclear power station.

View full abstract

Within the nuclear reactor safety analysis, one of the events that could potentially lead to a recriticality accident in case of a Small Break LOCA (SBLOCA) in a pressurized water reactor (PWR) is a boron dilution scenario followed by a coolant mixing transient. Some UPTF experiments can be interpreted as generic boron dilution experiments. In fact, the UPTF experiments were originally designed to conduct separate effects studies focused on multidimensional thermal hydraulic phenomena. But, in the case of experimental program TRAM, some studies are realized on the boron mixing: tests C3. Some of these tests have been used for the validation and assessment of the 3D module of CATHARE code. Results are very satisfying; CATHARE 3D code is able to reproduce correctly the main features of the UPTF TRAM C3 tests, the temperature mixing in the cold leg, the formation of a strong stratification in the upper downcomer, the perfect mixing temperature in the lower downcomer and the strong stratification in the lower plenum. These results are also compared with the CFX5 and TRIO-U codes results on these tests.

View full abstract

The Commissariat a L'Energie Atomique (CEA, French Atomic Energy Agency) and AREVA NC have developed new acidic surfactant liquid solutions to remove organic matter located at the surfaces of equipment used in reprocessing facilities. The aim of developing acidic formulations is to avoid sodium hydroxide to prevent uranium and plutonium oxides to precipitate and to ease the glass conditioning of the final liquid wastes. The organic matter providing the contamination is a solvent widely used in nuclear industry as a complexing agent of uranium and plutonium, the tributylphosphate (TBP). The purpose of this work is to understand and adapt the mechanisms involved in the TBP detachment and solubilization in acidic surfactant solution. In order to reach an optimal effectiveness, two well-known mechanisms should be combined: roll-up and emulsification. These mechanisms are characterized with a CCD camera allowing us to measure contact angles between a solid substrate and a liquid drop. For this work, we measured contact angles of a TBP drop deposited on a stainless steel plate, immersed in an acidic surfactant solution bath. In addition, we quantified the amount of TBP solubilized in the micelles by turbidity measurements. As a result, we formulated new acidic surfactant solutions with improvement factors in various fields (total organic matter amount, oil detachment and solubilization efficiency, emulsion stability…)

View full abstract