Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure by conducting probabilistic risk assessment for EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, human error probability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10 −6 /year in this paper. By considering the secondary sodium freezing, the fuel damage frequency was twice increased. The dominant accident sequence resulted from the common cause failure of the damper opening and/or the human error for the switching from the stand-by to the operation mode in the three stand-by cooling circuits. The second dominant accident sequence following the secondary pump trip is sodium freezing caused by the failure of air blower trip in the air cooler due to the common cause failures of secondary sodium flowmeter failure or erroneous opening of the air cooler damper. The Fussell-Vesely importance and risk achievement worth analyses have indicated high risk contributions.

Safety in nuclear facilities is of special importance and recent seismic events and the resulting damage in these facilities again bring up the discussion. One of the latest events is the 2007 Chuetsu earthquake in Japan. The arising damage in the Kashiwazaki Kariwa Nuclear Power Plant can be found in several reports like (Fukushima, 2007) and (Yamashita, 2008). Base-Control Systems (BCS) consist of Spring-damper combinations with specifically designed stiffness and damping properties which can be described in the vertical and both horizontal directions. Nowadays these 3 dimensional supports are frequently used to reduce seismic acceleration levels of machine components, equipment and buildings as safety or non-safety related structures. They may also be applied to protect sensitive equipment or machinery against other catastrophic events, i.e. airplane impact which is of particular importance in nuclear facilities. By presenting specific executed examples, this article shall also underline the efficiency of the BCS which, in contrary to purely horizontally efficient Base-isolation Systems with rubber (BIS), is not extensively discussed in literature. Decisive characteristics of the shown applications are provided and the benefits of using this technology for machine or equipment manufacturers, designers or owners are explained.

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K) [1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.

Control rod drive mechanism is very important to the operation and safety of reactors, therefore, requiring good mechanical properties and a high degree of reliability. The control rod drive mechanism of high-temperature-reactor pebble-bedmodule (HTR-PM) is powered by permanent magnet stepping motor and damped by the permanent magnetic damper. Therefore, the performance of both motor and damper have important impact on the control rod drive mechanism. However, the performance of permanent magnets is very sensitive to high-temperature. This paper describes the impact of the high temperature on the performance of critical components in the HTR-PM control rod drive mechanism. A Full-size thermal test bench is used to determine the performance of the key components within the scope from room temperature (20 ° C) to 250 ° C, with these data we can calculate the emergency insertion time. And further work plan is introduced too.

The prototype fast breeder reactor “Monju” has an ex-vessel fuel storage system (EVSS) which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST uses natural circulation of sodium for decay heat removal. Natural circulation in the EVST is generated by the decay heat from the spent fuel assemblies and the cooling of the cooling coils installed in the EVST. The EVST sodium cooling system consists of three independent loops. In each loop, sodium is circulated by electromagnetic pumps and the heat is removed by an air cooler with blowers. This system has the ability to remove the maximum decay heat using two loops, and thus, it uses two of the three loops for normal operation. During a station blackout (SBO), the pumps and blowers are stopped. However, the three air coolers are installed about 13.5 m higher than the cooling coils, and therefore, the EVST sodium cooling system potentially retains some cooling ability because of natural circulation. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed. The ultimate heat sink for the EVST sodium cooling system is the atmosphere, and the air coolers have an exhaust stack for efficient natural circulation caused by the chimney effect. However, the EVST sodium cooling system loses pressure and the heat transfer characteristics change if the flow rate is low. It was, therefore, necessary to confirm the temperature and flow rate behavior of EVSS in this analysis. In the present calculations, the plant dynamics analysis program “Super-COPD” was used. The factors affecting the cooling ability were investigated and analytical cases were determined. In one case, the two operated loops were switched to natural circulation after an SBO. The number of cooling loops was then changed from two to three by having an operator open the vane and dampers of the standby loop. In this case, sodium temperature in the EVST increased to approximately 320°C. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450°C. In both cases, however, the structural integrity of the EVSS was maintained. These analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.

One of the tests to assess the condition of generator stator end windings is the performance of a bump test. Bump testing is a form of nondestructive testing, as the stator end windings and similar components may be tested without harm of a mechanical or electrical nature. Bump testing analyzes the natural response of stator end-windings upon excitation from a mechanical impulse. The conductors, the bars and the complete stator end windings can be considered as a mechanical spring-mass-damper system, and therefore all have specific natural frequencies that depend on the rigidity, mass and damping. Since the windings and supports largely determine the rigidity, the height of the resonant frequency is dependent on the rigidity of the winding or support structure. Over time the insulation and supports can become loose and the natural frequency can become lower. With the results of bump testing, specific areas of concern can be accurately identified and the overall response plotted. Corrective actions can be taken to avoid an entire rewind of the generator stator, thereby saving considerable amounts of time and money.

Coupling structure interconnected by hysteretic dampers appears to be an effective method to mitigate structural seismic response. In the paper, the random seismic response is evaluated through the pseudo-excitation principle incorporated with stochastic equivalent linearization method without by solving the Lyapunov differential equation. For which, the seismic excitation is limited to be shot noise process and the computation burden should not be neglected while structural freedoms are large. In the paper, it is supposed that the structures keep elastic all the time and the hysteretic dampers are represented with versatile Bouc-Wen model. With the participation of assistant augment and reduced matrices which are correlated with the location of hysteretic dampers, the unidirectional excitation of one component and spatial excitation of multiple components are derived and the relationship between the pseudo-excitation and pseudo-response is deduced. Then, a pseudo-excitation closed-form expression for the system random response is established. Consequently, the stationary random seismic response of two shear type structure interconnected with hysteretic dampers is analyzed. The structural stationary seismic responses for two methods agree well. parametric studies for the hysteretic dampers and the optimum way to install the hysteretic dampers are also discussed.

The Fast Breeder Test Reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam, India. The reactor went critical in October, 1985 with a core of 23 unique high Plutonium carbide fuel subassemblies and the reactor power was rated for 10.5MWt with peak linear heat rating of fuel at 320W/cm. The extension of the target burn-up of this fuel based on Post Irradiation Examination at different stages enabled progressive expansion of the core and increase in reactor power. The reactor has been operated upto a power level of 18.6MWt/3MWe with a sodium temperature of 482°C max. The reactor has completed 24 years of operation and is currently under periodic safety review by the Atomic Energy Regulatory Board of India. As a part of the periodic safety review, equipment qualification status and ageing management studies have been presented to the regulators. Equipment qualification refers to the ability of the replaceable equipment to meet the functional requirements on demand, accomplished by periodic surveillance, maintenance and replacement. Ageing management addresses the residual life assessment of components which are passive, non-replaceable / replaceable with difficulty, taking into account their life degrading mechanisms. Over a period of time, based on the operational feedback, maintenance difficulties and obsolescence, several major components have been replaced. These include the Neutronic channels, UPS, computers of the Central Data Processing System, main boiler feed pumps, three control rod drive mechanisms, two control rods, central canal plug, deaerator lift pumps, reheaters of the steam water system, station batteries, DM plant and Nitrogen plant. The starting air system of the emergency diesel generators and isolation dampers of the reactor containment building have also been replaced. Regarding the non-replaceable components, residual life assessment has been carried out based on the operational history vis-a`-vis the design limits for each component. The life limiting mechanism of heat transport systems of FBTR are creep and fatigue. Since the reactor has operated only upto a temperature of 444°C till 2007, the creep effect is insignificant. The total number of thermal cycles seen by the reactor components as of 2007 was 163, as against the design cycle of 2000 for most of the components. Hence all the heat transport system components are as good as fresh ones. However, the major life limiting factor has been found to be the Neutronic fluence on the grid plate which supports the core. The fast flux at the grid plate location was measured using Np foils and the residual life of the reactor has been assessed to be 10.5 effective full power years. This paper details the life extension exercise being carried out for FBTR.

Nuclear power plant (NPP) design is strictly dependent on the seismic hazards and safety aspects related to the external events of the site. Passive vibration isolators are the most simple and reliable means to protect sensitive equipment from environmental shocks and vibrations. This paper concerns the methodological approach to treat isolation applied to a near term deployment reactor and its internals structures in order to attain a suitable decrease of response spectra at each floor along the height of the structure. The aim of this evaluation is to determine the seismic resistance capability of as-built structures systems and components in the event of the considered Safe Shutdown earthquake (SSE). The use of anti-seismic techniques, such as seismic isolation (SI) and passive energy dissipation, seems able to ensure the full integrity and operability of important structures and systems even in very severe seismic conditions. Therefore the seismic dynamic loadings, propagated up to the main reactor system and components, may be reduced by using the developed base-isolation system (high flexibility for horizontal motions) that might combine suitable dampers with the isolating components to support reactor structures and building. To investigate and analyze the effects of the mentioned earthquake on the considered reactor internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was used. To the purpose of this study a numerical assessment of dynamic structural response behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) were compared in order to highlight the isolation technique effectiveness.

To evaluate the ability of passive devices in protecting nuclear piping during earthquake a theoretical/experimental campaign has been performed. By means of numerical runs the effect of viscous dampers application on most critical points of a power plant steamline has been evaluated. The principle is to employ a local safety solution against heavy dynamic solicitations placing passive devices in crotch region of bends. The devices location corresponds to an in plane position in respect of the curve. Considerations on structural configuration and stress/strain states are also presented with the aim to respect the philosophy of design/verification requirements stated by the ASME Sct. III Cl.1 code. For experimental tests a C mock-up, whose sizes are derived by a thermal plant steamline, has been suggested and studied [1]. Comparison of numerical data on piping with/without dissipative elements are also included. The impact on the whole structure has been also taken into account. Some of the results included in the paper have been obtained in the E.U. contract named REEDS.

It has been established by other authors [1] that the accelerations of the water confined by the reactor core baffle plates has a significant effect on the responses of all the fuel assemblies during LOCA or seismic transients. This particular effect is a consequence of the water being essentially incompressible, and thus experiencing the same horizontal accelerations as the imposed baffle plate motions. These horizontal accelerations of the fluid induce lateral pressure gradients that cause horizontal buoyancy forces on any submerged structures. These forces are in the same direction as the baffle accelerations and, for certain frequencies at least, tend to reduce the relative displacements between the fuel and baffle plates. But there is another confinement effect — the imposed baffle plate velocities must also be transmitted to the water. If the fuel assembly grid strips are treated as simple hydro-foils, these horizontal velocity components change the fluid angle of attack on each strip, and thus may induce large horizontal lift forces on each grid in the same direction as the baffle plate velocity. There is a similar horizontal lift due to inclined flow over the rods when axial flow is present. These combined forces appear to always reduce the relative displacements between the fuel and baffle plates for any significant axial flow velocity. Modeling this effect is very simple. It was shown in previous papers [2,3] that the mechanism for the large fuel assembly damping due to axial flow may be the hydrodynamic forces on the grid strips, and that this is very well represented by discrete viscous dampers at each grid elevation. To include the imposed horizontal water velocity effects, on both the grids and rods, these dampers are simply attached to the baffle plate rather than “ground”. The large flow-induced damping really acts in a relative reference frame rather than an absolute or inertial reference frame, and thus it becomes a flow-induced coupling between the fuel and baffles. This has a significant effect on the fuel assembly motions and tends to reduce the relative displacements and impact forces between fuel assemblies and baffle walls.

The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology (Refs. 1 & 2) and RELAP5 (Ref. 3) to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL 96-06 methodology. This evaluation was based on a pressurized water reactor’s RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being reenergized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL 96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations and compares them.

The paper presents SERB–SITON method to control, limit and damp the shocks, vibration, impact load and seismic movements with applications in buildings, equipment and pipe networks (herein called: “components”). The elimination or reduction of shocks, vibration, impact load and seismic movements is a difficult problem, still improperly handled theoretically and practically because many times the phenomena are random in character and the behavior of components is non-linear with variations of the properties in time, variations that lead to the increase or decrease of the energy & impuls transfer from the dynamic excitation to the components. Moreover, the existing supports and dampers applied today, are not efficient enough in the reduction of the dynamic movement for all the frequency ranges met with in the technical application field. The stiffness and damping of classic supports do not allow a good isolation of components against shocks and vibrations so to eliminate their propagation to the environment and neither do they provide a satisfactory protection of the components sensitive to shocks and vibrations and seismic movements coming from the environment. In order to reduce the effects of shocks, vibrations impact and seismic movements on the components, this paper presents the results obtained by SITON on the concept, design, construction, experimental testing and application of new types of supports, devices and thin lattice structure, called “SERB”, capable to overtake large static loads, to allow displacements from impact, thermal expansions or yielding of supports and which, in any work position, can elastically overtake large dynamic loads or impact loads which they damp. The new supports and devices and thin lattice structure allow their adjustment without the occurrence of overstressing in the components due to their non–linear geometric behavior, and the contact pressure among the elements is limited to pre-set values to avoid blocking phenomena that generates great stresses induced by thermal expansion for example. Due to their characteristics of adjustment to the actual position and level of stress, SERB supports, devices and thin lattice structure show minimal effects on the components stress condition whenever the installation and computation errors. Herein below it is a presentation of the actual results obtained by SITON in the isolation of heavy equipment and pipe networks and others in process of application for buildings. Due to the very good results obtained in the isolation against shocks, vibrations and seismic movements at components in the conventional industry, there is the proposal to implement SERB-SITON method to the increase of the safety level at new or existing Nuclear Power Plants or to protect nuclear building against missiles and airplane crush impact.