Switchgear Operation and Maintenance for Process Plants (Process Plant Maintenance Book 1)

Construction shell type and core type Classification and type in relation to insulation, windings, core, cooling systems, voltage level, sizing, tank and breathing action Transformer parts. How to install, operate and work with high voltage power transformers safely Earthing of HV transformers. Surge protection Protective relaying differential, over-current and earth fault Buchholz relay and pressure relief relay Thermal devices and instruments oil temperature alarm and trip, winding temperature alarm and trip.

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Contact a Course Advisor about this Course! By checking this box, you agree to receive emails regarding your engineering career. Learning Online Remote Labs. Start your Application Today! To investigate the applicability of developed water radiolysis code, water chemistry at the water sampling point of the irradiation loop system was measured and compared with analytical results under several water chemistry conditions. Further, water chemistry distribution in the in-pile region as well as in the out-pile region was calculated by the developed water radiolysis code.

As a counter measurement of intergranular stress corrosion cracking IGSCC in boiling water reactors, the induction heating stress improvement IHSI has been developed as a method to improve the stress factor, especially residual stresses in affected areas of pipe joint welds. In this method, a pipe is heated from the outside by an induction coil and cooled from the inside with water simultaneously. By thermal stresses to produce a temperature differential between the inner and outer pipe surfaces, the residual stress inside the pipe is improved compression.

However IHSI had not been applied to weld joints of nickel-chromium-iron alloy P and austenitic stainless steel P Therefore for the purpose of applying IHSI to this one, we studied the following. The sputtering technique was used to protect the steel from corrosion. The thickness of sputtering-treated layer was All specimens were immersed in LBE in a pot for hours. The results showed that sputtering-treated layer still remained on the base of STBA No penetration of LBE was observed in this layer. The layer could protect the steel from penetration of LBE.

The result also showed that thin layer which contains aluminum oxide and chromium oxide was formed on the surface-treated layer, and it protected the base area. On the contrary, the penetration in base area was observed in the as received STBA Irradiation assisted stress corrosion cracking IASCC is one of the critical concerns when stainless steel components have been in service in light water reactors LWRs for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations PIEs.

Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. The tests were carried out in pure water simulated boiling water reactor BWR coolant condition. Stress corrosion cracking SCC is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques.

Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report. There are several types of strange corrosion phenomena from the point of view of our current understanding of corrosion science established in other fields. Through studying and coping with diverse corrosion phenomena, the author believes that they share a common basis with respect to the assumed corrosion mechanism e.

In general, local cell action is rarely severe since it produces a fairly uniform corrosion. If this mechanism is assumed in nuclear power plants, the structure becomes anodic in the area where the potential is less positive and cathodic where this potential is more positive. Metallic ions generated at anodic corrosion sites are transported to remote cathodic sites through the circulation of water and deposits as corrosion products.

This situation is the same as a battery that has been short-circuited at the terminals. No apparent external potential difference exists between the two electrodes, but an electrochemical reaction is still taking place inside the battery cell with a large internal short current.

In this example what is important is the potential difference between the local coolant and the surface of the structural material. It tends to be more hazardous because of its localized nature compared with the local cell action corrosion. There exist various mechanisms electrochemical cell configurations that induce such potential differences, including: In this paper, the author will discuss these potential differences and their relevance to the un-resolved corrosion issues in nuclear power plants.

Due to the importance of this potential mechanism the author is calling for further verification experiments as a joint international project. The analysis system consists of 3-D model established per the actual dimensions and interfaces of CRDM parts and a routine to calculate forces acting on the mechanism, and was verified by mock-up test using equipment same to actual product.

The analysis system is useful for functional evaluation in maintenance or to factor out root causes in case of malfunction of CRDM. An analytical study on micro-indentation method to integrity evaluation for graphite components was carried out. The indentation method is used as simplicity test to measure mechanical properties of materials. This method is thought to be applicable to evaluate the residual stress from the relationship between indentation load and indentation depth. Moreover, analytical investigations of indentation load-depth behavior for oxidized graphite and oxidized graphite with residual strain were also carried out.

As a result, it can be said that the indentation method is potentially applicable to evaluate the integrity of graphite components. The experimental study has been carried out to investigate reaction, transport and settling behavior of lead-bismuth eutectic LBE in flowing liquid sodium.

In the test, g of LBE were poured into flowing sodium from the top of a vertical-type sodium loop which contained The initial temperature of LBE and sodium was K. The sodium loop has a settling chamber at the lower part to investigate the concentration decrease behavior of solid particle reaction products in the sodium due to the settling effect. The concentration was measured by sodium sampling from the 11 positions of the loop during the experiment and its post-test chemical analysis.

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The temperature changes at the various parts of the loop were also measured during the experiment by thermo-couples attached on the outer surface of the loop. Ultrasonic detectors were attached on the outer surface of the loop below the position of a LBE pour nozzle to demonstrate the utility as a leak detector.

Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallisation effect of second cladding layer.

Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding. The strain rates were 0.

The dissolved oxygen concentration of the test water was maintained below 1 ppb. From our data, we confirm the occurrence of the dynamic strain aging DSA , and finally it can be considered that the primary hardening was brought about by the DSA. The secondary hardening was observed distinctly for 0. The improvement of fatigue resistance and the secondary hardening occurred under the same loading condition.

Therefore, the improvement of fatigue resistance may be related to the occurrence of the secondary hardening. When the secondary hardening occurs, intense slip bands are replaced by the corduroy structure. The corduroy structure can induce retardation of crack initiation, and ultimately the fatigue resistance is improved. Comparative study between the fatigue life generated in the current study and some prediction models was performed to evaluate the reliability of our data. Probabilistic analyses are interesting because some key variables, albeit conventionally taken at conservative values, can be modeled more accurately through statistical variability.

One variable which significantly affects RPV structural integrity assessment is cleavage fracture initiation toughness. However, in order to quantify the toughness scatter for probabilistic analyses, the master curve method is being analyzed at present. Furthermore, the master curve method is a direct means of evaluating fracture toughness based on K JC data. In the framework of the master curve investigation undertaken by EDF, this article deals with the following two statistical items: Concerning the first point, master curve temperature dependence is empirical in nature.

Our working hypothesis is that some ferritic steels may behave in slightly different ways. Therefore we focused exclusively on the basic french reactor vessel metal of types A Class 3 and A grade B Class 1, taking the sampling level and direction into account as well as the test specimen type. As for the second point, the emphasis is placed on the uncertainties in applying the master curve approach. For a toughness dataset based on different specimens of a single product, application of the master curve methodology requires the statistical estimation of one parameter: Because of the limited number of specimens, estimation of this temperature is uncertain.

The ASTM standard provides a rough evaluation of this statistical uncertainty through an approximate confidence interval. In this paper, a thorough study is carried out to build more meaningful confidence intervals for both mono-temperature and multi-temperature tests. These results ensure better control over uncertainty, and allow rigorous analysis of the impact of its influencing factors: A key problem in the development of heavy liquid metal cooled reactors is a corrosion of the structural and fuel cladding materials by the coolants.

Thus, the problem has been considered as an important design-factor that limits the operational temperature and flow velocity of the next generation nuclear reactors using lead-alloys. After each test, the specimens were analyzed metallurgically by using a scanning electron microscopy SEM with a energy dispersive X-ray analysis EDX for the cross sections of the specimens.

In addition, X-ray diffraction XRD was performed to evaluate the phase composition of the steels. This study couples structural analyses of four typical U. This method is used to determine the increase in the early release frequencies risk due to postulated cases of corrosion in the steel liners and shells, as well as other forms of degradation.

Environmental qualification testing was performed on a modified Limitorque torque switch for the torque switch safety functions in the Limitorque type SMB actuators located inside and outside containment in a nuclear power plant. The present paper describes the qualification testing performed. The modified torque switch was aged to a year service life at the normal service conditions for both inside and outside containment.

Aging included radiation, thermal and cycle aging. After each stage of aging, functional tests were done to confirm normal insulation resistance, normal contact resistance and normal operation. Pressure vessel integrity assessment after long-term service irradiation is commonly based on surveillance program results. Nevertheless, only the investigation of RPV material from decommissioned NPPs enables the evaluation of the real toughness response.

The operation of the four Greifswald units was finished in after 12—15 years of operation. In autumn the first trepans diameter mm were gained from the unit 1 of this NPP. Some details of the trepanning procedure will be given. The paper mainly deals with the retrospective dosimetry based on Niobium, which is a trace element of the RPV material. The neutron spectra were determined at the trepan positions. The different loading schemes of unit 1 standard and with 4 or 6 dummy assemblies were taken into account. The calculated specific 93m Nb activities for February, at the sample positions were determined to It turns out that the 93m Nb generation on the second path is nearly of the same order as the fast neutron induced generation from Niobium.

For the experimental determination of the 93m Nb-activity the Nb-content was determined by ICP-MS inductive coupled plasma mass spectrometry after dissolution of the material sample. The radiochemical isolation of Nb was done by anion exchange separation. The radiochemical separation was accompanied by determination of the chemical yield of Nb using again the ICP-MS method.

Possible reasons for the observed differences are discussed. New plants are on the horizon! Several utilities have announced intentions of breaking ground in , meaning that heavy component orders will have to be placed as early as ! With an average lead time of 10 years, the bulk of the new plant orders will have to be placed by not later than This equates to almost 8—9 new plant orders a year!!

KAERI has also investigated the conceptual design of a lead-cooled fast reactor. Lead Pb is used as the coolant material in that reactor. The most significant problem is a corrosion when using the lead-alloy liquid metal in those reactors. Therefore, it is necessary to study the corrosion characteristics and develop the technology to protect the steel structure materials against a corrosion.

KAERI has been developing the facilities needed to study the corrosion of lead-alloy. KAERI fabricated a static corrosion test facility in The Pb-Bi loop was constructed in The first stage of the Pb loop construction was finished and operations began in We will complete the second stage of the Pb loop construction after testing the first stage Pb loop. The complete understanding of the incubation and growth of microstructurally short cracks is still somewhat beyond the present state-of-the-art explanations.

A good example is the intergranular stress corrosion cracking of Inconel in high-temperature water. An effort was therefore made by the authors to construct a computational model of the crack growth kinetics at the grain-size scale. The main idea is to divide continuum e.

Random grain structure is modelled using Voronoi-Dirichlet tessellation. Each grain is assumed to be a monocrystal with random orientation of the crystal lattice. Elastic behaviour of grains is assumed to be anisotropic. Crystal plasticity is used to describe small to moderate plastic deformation of monocrystal grains. Explicit geometrical modelling of grain boundaries and triple points allows for the development of the incompatible strains along the grain boundaries and at triple points. The analysis is currently limited to two-dimensional models. Numerical examples illustrate analysis of about one grain boundary long transgranular cracks.

In particular, the dependence of crack tip displacements on the random orientation of neighbouring grains is studied. The limited number of calculations performed indicates that the incompatibility strains, which develop along the boundaries of randomly oriented grains, significantly influence the local stress fields and therefore also the crack tip displacements.

First attempts are also made to quantify the preferential growth directions of cracks crossing the discontinuities e. Exposure of concrete to elevated temperature affects its mechanical and physical properties. Elements could distort and displace, and, under certain conditions, the concrete surfaces could spall due to the buildup of steam pressure. Because thermally-induced dimensional changes, loss of structural integrity, and release of moisture and gases resulting from the migration of free water could adversely affect plant operations and safety, a complete understanding of the behavior of concrete under long-term elevated-temperature exposure as well as both during and after a thermal excursion resulting from a postulated design-basis accident condition is essential for reliable design evaluations and assessments of nuclear power plant structures.

As the properties of concrete change with respect to time and the environment to which it is exposed, an assessment of the effects of concrete aging is also important in performing safety evaluations. The effects of elevated temperature on Portland cement concretes and constituent materials are summarized, design codes and standards identified, and considerations for elevated temperature service noted.

All over the world numerous fatalities and large property losses have taken place because of Industrial Explosions. In order to mitigate the losses of life and property due to overpressure rupture discs are used. Successful implementation would depend on many factors and one of the important factors is the selection and design of rupture discs. There are several factors that affect the disc performance. The present paper discusses the factors that effect the disc performance so that the right type of disc can be selected. The author has investigated the characteristics of boron co-deposition with crud experienced in AOA and iron ferrite deposition in CDA.

Corrosion product deposits found in cores with appreciable AOA have been reported in mostly nickel-based as NiO or elemental nickel as opposed to nickel ferrite deposits common to non-boiling cores. Although preliminary, the estimated E-pH diagram suggests some interesting observation, including: During the AOA cycle, the concentration of nickel and iron ions must have been unusually high as they should be for a significant amount of crud deposits.

The author thinks such an acceleration of the anodic dissolution of metal cations is due to the effect of the long cell action corrosion mechanism. As early as , an Italian scientist Petracchi demonstrated that electrochemical effects significantly influence the erosion rate. He constructed a flow nozzle with specimens kept under external electrical potential. Upon inducing as low as 0. No erosion was observed by reversing the polarity of the potential.

As discussed in a companion paper also presented at this conference [1], the author discusses various mechanisms electrochemical cell configurations that induce potential differences, including those differences in ionic concentration, aeration, temperature, flow velocity, radiation and corrosion potentials. The author is calling for further verification experiments regarding this corrosion mechanism as a joint international project. Operating cycles are nominally seven per year with outages that last 7 to 14 days, allowing time for routine plant maintenance and experiment insertions and manipulations.

While the ATR pressurized water loops can operate at the same temperature and pressure requirements of a pressurized water reactor, the loops also have the ability to operate at higher conditions. Hence, it is critical to ensure that when component replacements are called for, they can meet or exceed design requirements of a typical power reactor, while continuing to satisfy the design requirements of the ATR experiment loops. Irradiation of materials by energetic particles causes significant degradation of the mechanical properties, most notably an increased yield stress and decrease ductility, thus limiting lifetime of materials used in nuclear reactors.

The microstructure of irradiated materials evolves over a wide range of length and time scales, making radiation damage and inherently multi-scale phenomenon. At atomic length scale, the principal sources of radiation damage are the primary knock-on atoms that recoil under collision from energetic particles such as neutrons or ions.

These knock-on atoms in turn produce vacancies and self-interstitial atoms, and stacking fault tetrahedra. At higher length scale, these defect clusters form loops around existing dislocations, leading to their decoration and immobilization, which ultimately leads to radiation hardening in most of the materials. All these defects finally effect the macroscopic mechanical and other properties. An attempt is made to understand these phenomena using molecular dynamics studies and discrete dislocation dynamics modelling.

The combination of one or several processes of ruins can involve the materials failure of a nuclear power plant. The prolongation and the repetition of these effects can involve a deterioration of the machine. In accordance with the decree of February 26, , the PWR operator must be firstly, sure that the system is controlled according to the situations considered in the file of dimensioning and secondly, be able to know anytime the life of the equipment. In the SFR, the high temperatures imposed on components for long periods can involve a significant creep. In the course of time, this deformations accelerate the release of fatigue cracks.

To consider the creep, the reactor lifespan is correlated at the numbers of thermals transients envisaged initially. To realize the management of aging in Phenix power plant, it is necessary to carry out an individualized monitoring of the structures and not only on the vessel. A profound knowledge of the thermal transients of the past is necessary to carry out an effective assessment.

In order to guarantee that any harmful situation is well taken into the management of aging, we monitor permanently certain measurements primary and secondary pump speed, hot and cold pool temperatures, IHX-main vessel and reactor roof temperatures. We present in the article the scientific method used in the Physics Section. A logical diagram specific to the type of situation and the structure allows to associate the harmful transient at a identical situation which has been happened in the past.

During the last two cycles, the nuclear power plant has sustained 34 startup 20 during the 51st cycle and 14 during the 52nd cycle. For the last 4 cycles, the number of transients to come will remain quite lower than the number dimensioned initially. SA type LN stainless steel material, having closer control over impurities and inclusion content, is the intended piping material in the Advanced Heavy Water Reactors.

Deformation, fatigue and fracture behaviour of this material and its weldments have been characterized at ambient temperature and at K. The details of the fractographic investigations and stretch zone width measurements are also discussed. This is attributed to larger density of second phase inclusions in the SMAW weld metal.

SZW measurements give a good alternate estimate of the toughness of the materials. Fatigue crack growth rate in SMAW weld metal was found to be comparable to base metal at higher load ratios. For the development of the System Based Code, which was proposed by Asada and intends to optimize structural design of nuclear components by enabling margin exchange between various technical options, a tool for life cycle structural reliability evaluation method is necessary.

The system is primarily for fast breeder reactors but its methodologies can also be applied to the other types of reactors. A matrix containing inert gas bubbles dilates in direct proportion to the growth experienced by the gas bubbles. This phenomenon is termed as swelling. A model for the swelling induced by the growth of the helium gas bubbles in irradiated copper-boron alloys is presented. The bubbles grow by acquiring vacancies from the external surface, which acts as a source of vacancies.

The vacancies reach the surface of the bubbles mainly via lattice diffusion and to a limited extent via diffusion through short-circuiting paths such as grain boundaries and dislocation pipes. The model predicts that overall swelling of the matrix varies as 1. Another consequence of the present model is that the growth rate of a gas bubble varies inversely as the cube of its distance from the external surface. The model has been applied to the data on irradiated copper-boron alloys and found to be in accord with the experimental results. The model is general and can be applied to the growth of all kinds of stationary inert gas bubbles trapped within a crystalline matrix.

This paper provides information on swing check valve selection criteria suitable for nuclear power plant applications. In this project, four swing check valves were analyzed to demonstrate the implementation and application of this information. The use of CFD is a relatively new approach for validation of valve design that is becoming invaluable due to the high cost of physical bench testing. Today, there are several units in operation at the Pickering, Bruce, and Darlington sites in Ontario, Canada. The steam generator tubing materials were manufactured from Monel , Inconel , and Incoloy for the Pickering, Bruce, and Darlington respectively and are subjected to different operating conditions.

This paper presents a review of some of the various types of degradation mechanisms that have been observed on these tubing materials over the operating period of the respective plants. The results presented are based on the metallurgical examination of removed tubes. The mechanisms that have been observed include pitting, stress corrosion cracking, intergranular attack, fretting, and erosion corrosion. The nature of the flaws and causative factors if known are discussed. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors.

This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort.

The late Professor Emeritus Yasuhide Asada proposed the System Based Code concept, which intends the optimization of design of nuclear plants through margin exchange among a variety of technical options which are not allowed by current codes and standards. The key technology of the System Based Code is margin exchange evaluation methodology. This paper describes recent progress with regards to margin exchange methodologies in Japan. The key concepts of the System Based Code are; 1 life-cycle margin optimization, 2 expansion of technical options as well as combinations of technical options beyond the current codes and standards, and 3 designing to clearly defined target reliabilities.

Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to.

Plant Equipment and Maintenance Engineering Handbook

On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept.

It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper.

On December , the French regulator issued a new regulation for French nuclear power plants, in particular for pressure equipment PE. This regulation need first to agree with non-nuclear PE regulation and add to that some specific requirements, in particular radiation protection requirements. Only few components are nuclear specific. In parallel to that, a new safety classification has been developed by French utility. The consequences is the need to cross all these specifications to define a minimum quality level for each components or systems.

This paper will summarize the key aspects of these different topics. Some of these JSME nuclear codes have been endorsed by the regulatory body, and are now utilized in the regulatory processes of the actual plants. It includes as its main body design rules on class 1 components for elevated temperature services. This paper overview the main features of the code.

Comprehensive reformation of the regulatory system has been introduced in Japan in order to apply recent technical progress in a timely manner. The revised Ordinance No. During the period from the issuance to the enforcement, JNES carried out to prepare enforceable regulatory guide which complies with each provisions of the Ordinance No. Some consensus codes and standards were re-assessed since they were already used in regulatory review of the construction plan submitted by licensee.

Other consensus codes and standards were newly assessed for endorsement. In case that proper consensus code or standards were not prepared, details of regulatory requirements were described in the regulatory guide as immediate measures. At the same time, appropriate standards developing bodies were requested to prepare those consensus code or standards. Supplementary note which provides background information on the modification, applicable examples etc. Although the government admit the benefit of construction of a nuclear facility for national electric source, related policy could be developed and carried out only if the public, especially who have some stake on it, recognize the benefit and accept the policy.

For public participation, Korea has a system of public-hearing in accordance with the law. Because of the absence of the detailed way for public opinion aggregation and for the reflection of the aggregated opinion, Korean public-hearing system is only a conceptual model. Therefore, some specific system for Korean Public-Hearing should be developed and applied.

In this study, to share the right of decision making, which is an ultimate concept for public participation, decision making components and the characteristics of each phase are analyzed. The criteria weight for assessment and comparison with alternatives are founded as a valuation factor of the decision making components, which should be based on the social consensus. On these foundations, a system for aggregation and reflection of the public opinion was proposed. Also, in this study, a mathematical model for the quantification of the degree of consensus was conceptualized using Ordered Weighted Averaging OWA aggregation operator and fuzzy similarity theory, which is a comparison concept.

Since this model enables influence of each criteria and each participant on collective consensus to be analyzed, a direction to promote consensus building can be derived. That is to say, this model can support consensus building and promote public acceptance for the nuclear industry and related policy. Risk-informed inservice inspection ISI programs have been in use for over seven years as an alternative to current regulatory requirements in the development and implementation of ISI programs for nuclear plant piping systems. Additionally, many plants have conducted or are in the process of conducting updates to their risk-informed ISI programs.

In the development and implementation of these risk-informed ISI programs and the associated updates to those programs, the following important lessons learned have been identified and are addressed. The insights gained are associated with many of the steps in the risk-informed ISI process including: Many of these lessons learned have impacted the results of the risk-informed ISI programs and have impacted the updates to those programs.

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With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world such as United States, France, Japan, Canada, South Korea, South Africa and they will be marketing their reactors to many countries around the world such as US, China, South Korea, France, Canada, Finland, Taiwan. They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world.

To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. Silicon carbide SiC based uranium ceramic material can be fabricated as hosts for ultra high temperature applications, such as gas-cooled fast reactor fuels and in-core materials. A pyrolysis-based material processing technique allows for the fabrication of SiC based uranium ceramic materials at a lower temperature compared to sintering route.

Modeling of the process is considered important for optimizing the fabrication and producing material with high uniformity. This study presents a process model describing polymer pyrolysis and uranium ceramic material processing, including heat transfer, polymer pyrolysis, SiC crystallization, chemical reactions, and species transport of a porous uranium oxide mixed polymer. Three key reactions for polymer pyrolysis and one key reaction for uranium oxide polymer interaction are established for the processing. Included in the model formulation are the effects of transport processes such as heat-up, polymer decomposition, and volatiles escape.

The model is capable of accurately predicting the polymer pyrolysis and chemical reactions of the source material. Processing of a sample with certain geometry is simulated. The effects of heating rate, particle size and volume ratio of uranium oxide and polymer on porosity evolution, species uniformity, reaction rate are investigated.

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. A comparison of the code predictions vs.

There have been a lot of tests and analyses reported for evaluation of drop tests of metal casks. However, no quantitative measurement has ever been made for any instantaneous leakage through metal gaskets during the drop tests due to loosening of the bolts in the containments and lateral sliding of the lids. In order to determine a source term for radiation exposure dose assessment, it is necessary to obtain fundamental data of instantaneous leakage. In this study, leak tests were performed by using scale models of the lid structure and a full scale cask without impact limiters simulating drop accidents in a storage facility, with aim of measuring and evaluating any instantaneous leakage at drop impact.

Prior to drop tests of a full scale metal cask, a series of leakage tests using scale models were carried out to establish the measurement method and to examine a relationship between the amount of the lateral sliding of the lid and the leak rate. It was determined that the leak rate did not depend on the lateral sliding speeds. Drop tests of a full scale metal cask without impact limiters were carried out by simulating drop accidents during handling in a storage facility.

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The target was designed to simulate a reinforced concrete floor in the facility. The first test was a horizontal drop from a height of 1 m. The second test simulated a rotational impact around an axis of a lower trunnion of the cask from the horizontal status at a height of 1 m. In the horizontal drop test, the amount of helium gas leakage was calculated by integrating the leak rate with time.

The total amount of helium gas leakage from the primary and secondary lids was 1. This value is 9. The amount of leakage was insignificant. In the rotational drop test, the total amount of leakage from the primary and secondary lids was 1.

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This value is 8. This value was larger than that of the horizontal drop test. Nevertheless, the amount of leakage was also insignificant. The relationship between the maximum sliding displacement of the lid and the leak rate coincided between the tests of a scale model and a full scale metal cask.

ABB vacuum circuit breaker switchgear VD4 maintenance cement plant part2

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined by flowsheet analysis such that the satisfactory decontamination is achieved.

Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced.

The target system, whose function is to supply an external neutron source to the ADS sub-critical core to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due to accelerator protons and fission neutrons. This document deals with the thermo-hydraulic results of the calculations performed with STAR-CD and RELAP5 codes for studying the behaviour of the windowless target system during off-normal operating conditions. It also reports a description of modifications properly implemented in the codes needed for this analysis.

The windowless target system shows a satisfactory thermo-hydraulic behaviour for the analysed accidents, except for the loss of both pumps without proton beam shut-off and the beam trips lasting more than one second. Afterwards, performances comparisons are made in terms of maximum deployable power, natural uranium consumption and waste production. The results show that the FBR maximum deployable capacity, independently from the FBR technology, is highly sensitive to the fuel cycle options, like the spent nuclear fuel cooling time or the Minor Actinides management strategy.

Crystallization is one of the remarkable technologies for future fuel reprocessing process that has safety and economical advantages.

Plant Equipment and Maintenance Engineering Handbook

In previous study, we carried out experimental studies with uranium, MOX and spent fuel conditions, and flowsheet analysis was considered. From the design study, an annular type continuous crystallizer was selected as the most promising design, and performance was confirmed by small-scale test and engineering scale demonstration at uranium crystallization conditions. In this paper, the design study and the demonstration test results are described. Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing.

There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack.

Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking SCC of the canister is not sufficient.

Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility.

Air including the sea salt particles goes into the dry storage facility Figure 1. Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding.

The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building.

For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack. This code has been designed to study short, medium and long term options for the introduction of various types of nuclear reactors and for the usage of associated nuclear materials.

It permits to study transition scenarios and gives due consideration to isotopic composition essentially of uranium, plutonium, minor actinides and some fission products. In the frame of the French law for the researches about waste management, different dynamic scenarios have been studied [1]. These scenarios are considering the French case and start from the present situation, which consists in a single stage of Plutonium recycling in PWRs.

The scenarios described in this paper take into account two main options: Continuation of nuclear energy or phase out option. This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels SNF. It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned.

This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report. The nuclear spent fuel transport and storage cask is used for transport of the spent fuel from a nuclear power station to an intermediate storage facility.

Leak tightness and subcriticality on transportation required from IAEA TS-R1 [1] have to be assured by a 9m drop test and its numerical simulation. This paper describes the drop test using a full-scale prototype test cask. The drop orientations of the tests were slap down and vertical. From the drop test the following is demonstrated: This means the simulation method qualified the relation of dynamic response of the cask body and leakage behavior.

In this study, we have proposed a new disassembly technology by mechanical disassembly system that consists of a mechanical cutting step and a wrapper tube pulling step. The fuel pins are transported to the shearing device in next process. The Fundamental tests were carried out with simulated FBR fuel pins and wrapper tube, and cutting performance and wrapper tube pulling performance has been confirmed by engineering scale. As results, we established an efficient disassembly procedure and the fundamental design of mechanical disassembly system.

An existing melter is the second melter, which was installed from to in place of the first melter stopped its operation by damage of a main electrode. JAEA has estimated that the damage was caused by accumulation of noble metal. Therefore, melter bottom structure was improved to get better drain ability of glass containing noble metal.

Completing the melter replacement, vitrification operation was restarted in October and produced 88 canisters successfully until the end of March Based on the basic strategy, JAEA made a decade development plan of necessary key technologies and has started the development since Burnup limitations are normally set to limit stresses in the fuel assembly components.

The defined limits provide guidance to the fuel designer to minimize fuel failure during steady sate operation, and also prevent against some thermal and mechanical phenomena that could occur during overpower transients. In particular, a LHGR limit value is set to take into account physical phenomena that could lead to pellet-cladding interaction.

This limit value directly relates to a PCI limit, which may be set based on experimental ramp tests. Thus, to avoid violating the PCI limit, fuel conditioning procedures are still required for both barrier and non-barrier fuel. Simulation of the power ramp procedures to be performed by the reactor operator during startup or power increase maneuvers is advisable as a preventive measure of possible overpower consequences on the fuel thermomechanical behavior. In this paper, the thermomechanical behavior of two different kinds of BWR fuel rods is analyzed for fuel preconditioning procedures.

Five different preconditioning computations were performed, each with three different ascending linear power rate ramps.