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Objectives This study focuses on the sealing performance of lifting hatches applied to nuclear power plants under accidental working conditions.MethodsAn experimental setup is built to achieve the quantitative measurement of steam micro-leakage. In this device, saturated steam at 0.66 MPa is charged into a pressurized chamber to simulate accident conditions, the leaking steam is collected through a collection chamber with an insulation layer, and no condensation occurs. The changes in temperature, humidity and pressure in the collection chamber are monitored in real time to estimate the steam leakage rate.ResultsThe experimental results show that the steam leakage rates of the lifting hatch models with 1.5 mm and 3.0 mm composite gaskets are 13.93 g/24h and 11.91 g/24h respectively, both meeting the low leakage design requirements, with the 3.0 mm gasket providing superior sealing performance. An analysis of the dry air and steam measurements with the same gasket and at the same pressure shows that the sealing coefficients nd4 obtained for both are very close, which verifies the accuracy of the experimental device for steam microleakage measurements.ConclusionsThe proposed steam microleakage measurement method can be extended to various sealing structures and pressure systems, and provides a new measurement idea for evaluating the sealability of stack chambers under accident conditions.

Changes in temperature during reactor operation may cause changes in physical parameters, leading to core overheating and accidents. It is essential to analyze and assess the safety parameters of the core under different operating conditions. This paper investigates the effects of fuel temperature, moderator density, boron concentration, and control rod state on AP1000 safety parameters. The study uses RMC and NJOY to calculate the changes in reactivity factor, effective delayed neutron fraction, and neutron generation time of the AP1000 reactor under different operating conditions. The changes in reactivity coefficients, neutron fluxes, and relative power densities of AP1000 reactors are analyzed for normal and accidental operating conditions. The results indicated that the reactivity coefficient remained negative under accident conditions, which ensured the safe operation of the reactor. The delayed neutron fraction, neutron flux, and power density distributions are affected by fuel temperature, moderator density, and control rod position. The control rod worth was sufficient for the emergency shutdown of the reactor under accidental conditions. It is demonstrated that the operation of the AP1000 reactor under study conditions is safe and controllable. [ABSTRACT FROM AUTHOR]

Changes in temperature during reactor operation may cause changes in physical parameters, leading to core overheating and accidents. It is essential to analyze and assess the safety parameters of the core under different operating conditions. This paper investigates the effects of fuel temperature, moderator density, boron concentration, and control rod state on AP1000 safety parameters. The study uses RMC and NJOY to calculate the changes in reactivity factor, effective delayed neutron fraction, and neutron generation time of the AP1000 reactor under different operating conditions. The changes in reactivity coefficients, neutron fluxes, and relative power densities of AP1000 reactors are analyzed for normal and accidental operating conditions. The results indicated that the reactivity coefficient remained negative under accident conditions, which ensured the safe operation of the reactor. The delayed neutron fraction, neutron flux, and power density distributions are affected by fuel temperature, moderator density, and control rod position. The control rod worth was sufficient for the emergency shutdown of the reactor under accidental conditions. It is demonstrated that the operation of the AP1000 reactor under study conditions is safe and controllable.

The dense granular flow spallation target is a new target concept proposed for an Accelerator-Driven Sub-critical (ADS) system. In this paper, the beam-target configurations of a tungsten granular flow target for the ADS with a thermal power of 1 GW is explored. The beam profile options using different scanning methods are discussed. The critical geometry parameters are adjusted to investigate the performance of the granular target from the aspects of neutron efficiency, stability and temperature distribution in target medium. To figure out how the target under accident conditions would behave, different clogging conditions are induced in the simulation. The dynamic processes are analyzed and some important parameters such as abnormal temperature rise and beam cutoff time window are obtained. The response of the sub-critical reactor to a clogging accident is also investigated. It is indicated that the monitoring of the granular flow by the neutron detectors in the sub-critical core will be effective.

This essay focus on the situation when the operator of the main control room can not handle the accident in time or make wrong judgment under accident condition in nuclear power plant, which may lead to the occurrence of major accidents or even more accidents concurrently. An intelligent fuzzy diagnosis model based on Grey Relational degree is constructed in this paper, by establishing accident-state correlation matrix and accident-state probability matrix, defining the vector of information weights for accident characteristic vector and monitoring result vector with linear transform, using Grey Relation to measure the relevance degree between monitoring result vector and accident characteristic vector, Searching for suspected accidents approaching to real solutions in iterative recursive algorithms, realizing automatic fuzzy diagnosis of multiple accidents. The results of simulation experiments of LOCA accidents indicate that the model and algorithm can diagnose various common accidents accurately and rapidly in complex nuclear power systems, providing strategy for operator's diagnosis decision. [ABSTRACT FROM AUTHOR]

In case of a temporal undefined failure of the active cooling system of a spent fuel pool (SFP), the water inside the pool heats up depending on the decay heat of the stored spent fuel. In the worst case, this leads to a boil-off scenario with a water level decrease and, after appropriate time, to a progressive uncovery of the fuel assemblies (FA). The cooling of the uncovered structures is limited by the steam production in the lower still wetted part of the FA, radiative heat transfer, radial heat conduction and convection inside the FA and interaction with ambient air depending on the storage conditions. As a result, the removal of decay heat can be endangered and, by reaching certain cladding temperatures, the integrity of the cladding tubes can be compromised. The detailed heat and mass transfer mechanisms of the boil-off scenario are not yet sufficiently investigated – especially experimentally. Therefore, reliable conclusions about the development of the cladding temperatures cannot be made. Against this background, the SINABEL project was launched in 2013 at TU Dresden for an improved understanding of the heat transport processes inside a SFP under accident conditions. Studies are performed both experimentally on the test facility ALADIN and theoretically by using computational fluid dynamics (CFD) calculations and integral codes for severe accidents. The test facility simulates a generic boiling water reactor (BWR) FA on almost original scale with regard to the boundary and ambient conditions within a SFP. Through an appropriate unique net of instrumentation, the distribution and development of radial and axial cladding temperatures and the development of the water level can be investigated in detail. Experiments were carried out for various decay heats respectively rod powers and changing storage distances. The thermal hydraulics of the heat-up and the boil-off phase with the occurrence of subcooled boiling in the beginning followed by geysering-like instabilities are described in the present paper. The higher the heat load, the faster the heat-up and, as boiling temperature is reached, the lowering of the water level. The maximum limit of the water level and the influence by the radial heat conduction, convection and radiative heat transfer is shown and explained. The experimental data and thermohydraulic findings provide the basis for the phenomenological understanding of this accident scenario and are used for the validation and improvement of numerical simulations. [ABSTRACT FROM AUTHOR]

U 3 Si 2 fuel and FeCrAl cladding have been reported as candidates of ATF materials. Efforts have been made to study the fuel performance of the U 3 Si 2 –FeCrAl system under normal operating conditions, and its fuel performance is also investigated under accident conditions. In this work, the fuel performance of U 3 Si 2 –FeCrAl system is investigated and compared to those of U 3 Si 2 -Zircaloy, UO 2 –FeCrAl, and UO 2 -Zircaloy systems under LOCA and RIA conditions. And a series of simulations of temperature evolutions with different heat flux are conducted to explore the effect of the heat flux on the temperature. Additionally, sensitivity of the fuel input parameters is analyzed to reveal the main impact on the fuel performance under LOCA condition. Due to the higher thermal diffusivity and lower capacity of the U 3 Si 2 fuel under LOCA condition, the performance of U 3 Si 2 fuel is found to be inferior to UO 2 fuel with the same cladding, and the heat flux of cladding outer surface is found to influence the temperature correlations between different fuels. Furthermore, the study of U 3 Si 2 fuel and FeCrAl cladding under RIA condition is also conducted due to their low melting point, and the U 3 Si 2 fuel is found to exceed their melting point in an early time under RIA condition. [ABSTRACT FROM AUTHOR]

The internal and external cooled annular fuel was proposed firstly by MIT, which is found to be able to improve fuel performance dramatically. Based on the CAMPUS code, an annular fuel performance code called CAMPUS-ANNULAR is further developed in this work. And the multiphysics models of annular UO 2 fuel equipped with Zircaloy cladding are developed under reactor normal operating, Loss Of Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA) conditions, respectively. Then the BISON cladding failure criterion utilized under LOCA condition is introduced, and a loss of coolant and reflood model is implemented under LOCA condition for the first time. Additionally, the radial power distribution of annular fuel is introduced and analyzed in this work. Finally, the models implemented in this work are verified and the annular fuel performance under normal operating, LOCA and RIA conditions are investigated and compared with those of solid fuel. The fuel maximum temperature is found to be reduced dramatically by using annular fuel under normal operating, LOCA and RIA conditions, which may further lead to the cladding failure time to be delayed significantly under LOCA condition. Besides, the risk of fuel melting is found to be reduced by using annular fuel under RIA condition. • CAMPUS-ANNULAR code is developed to study annular fuel performance. • The loss of coolant and reflooding model is firstly implemented to give more accurate analysis under LOCA condition. • The safety margin under both normal operating and accident conditions can be improved by using annular fuel. [ABSTRACT FROM AUTHOR]

The evaluation of fluid flow rates through crack-like leaks in pressurized components of NPPs plays an important role for leak-before-break considerations. When using available simplified methods for the estimation of critical discharge flow rates, a suitable friction factor has to be chosen. Guidelines are given in the German code KTA 3206 based on investigations of leak rate experiments summarized in this paper. The frictional pressure loss in different leak flow approaches has to be chosen consistently in order to obtain comparable results with different models. Thermal hydraulic and structural mechanical analyses were performed for a postulated leak in the pressurizer surge line (SL) of a PWR Konvoi type under design basis accident conditions. The leak was postulated in form of a through-wall crack in circumferential direction. The size of the leak was calculated in the framework of Finite Element calculations with the code ADINA using an analysis model of a cooling loop of Konvoi type. With the calculated leak size, an ATHLET calculation was conducted, also examining the influence of the consideration of a variable leak size depending on the system pressure and mechanical loads. In the investigated case, the implication of the decreasing leak size, especially on the course of the remaining system pressure, cannot be neglected. The reduction of the leak area amounts in the transient examined to ca. 25% after about 1 h due to both the system pressure and temperature decrease and leads therefore also to an approximately 23% smaller leak rate. These results are dependent on the assumed position of the leak. Concerning the determination of the leak rates as critical flow-through rates with simplified methods good agreement has been achieved although the approaches differ considerably in parts. Relevant influence factors are the inflow losses in the crack channel, the consideration of the hydraulic diameter in connection with the leak area and the assumptions on the flow resistance coefficient due to the roughness of the crack surfaces. [ABSTRACT FROM AUTHOR]

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel. [ABSTRACT FROM AUTHOR]

• Models of thorium-based fuels and two-layer SiC have been implemented into codes. • Thorium-based fuels with two-layer SiC cladding show improved fuel performance. • Thorium-based fuels with two-layer SiC reduce the cladding deformation under LOCA. Due to the excellent properties of high temperature strength, creep resistance, lower thermal expansion and better irradiation resistance, the silicon carbide is considered as a substitute for the Zircalloy cladding. In this paper, the thermal and mechanical properties of SiC cladding are reviewed firstly. Then the thermophysical properties and mechanical behavior model of the thorium-based fuel and SiC cladding are implemented in FRAPCON-4.0 and FRAPTRAN-1.5 codes. Lastly, the fuel performance of five different types of fuel cladding combinations including two kinds of thorium-based fuel, traditional UO 2 fuel and two kinds of cladding (Zircalloy and two-layer SiC) are analyzed and compared both under the PWR normal operating condition and transient conditions. The results show that under normal or transient conditions, the thorium-based fuel with two-layer SiC cladding is found to have higher safety margin and superior fuel performance than that of the fuels combined with Zircalloy cladding. [ABSTRACT FROM AUTHOR]

During the 1950s to the 1970s, when many new packages were being designed for the transport of a wide variety of radioactive materials, methods to analyse the response of package structures to structural accident environments had not been well developed. During this period, a number of different package designs were tested to determine how they reacted to various structural insults. Test programmes, on obsolete and new casks, were undertaken on behalf of the US Department of Energy and the Japanese Power Reactor and Nuclear Fuel Development Corporation (currently known as the Japan Atomic Energy Agency). At the Oak Ridge National Laboratory (ORNL), studies focused on how such packages reacted to the structural test environments specified in the transport regulations. At Sandia National Laboratories (SNL), studies focused primarily on tests that went beyond those required by the regulations. Two of the cask designs tested at ORNL were also tested to extreme environments at SNL, which allowed direct comparison of results for the different test environments. In addition, SNL exposed other casks to test conditions that were more extreme than were required by the regulations. Unfortunately, much of the original documentation regarding these tests and their results is no longer readily available or is degrading with time. The documentation relating to these tests and their results have been surveyed; this paper presents a summary from this survey. The first part of this paper describes the structural testing of three casks in accordance with regulatory standards, two of which were obsolete casks, the third of which was a newly designed cask for the transport of fresh nuclear fuel. The latter testing subjected the cask to 9 m (30 ft) drop and 1 m (40 in) puncture and a 258 m s–1 (848 ft s–1) impact test of the truck system into a massive concrete barrier and permitted a direct comparison of the results. The second part of this report discusses structural testing of casks to various structural environments different from those specified in the Transport Regulations. Where possible, the results of the regulatory tests are compared with the results of the tests to the other environments. [ABSTRACT FROM AUTHOR]

Beginning as early as the 1960s, concerns were voiced as to the adequacy of the package test standards imposed by the transport regulations promulgated by the International Atomic Energy Agency. One concern that was frequently raised and has continued to the present time is that the test standards do not necessarily simulate real accidents. The purpose of the crash tests described here that were done with typical packages carried in full scale vehicles was to assess the IAEA standards, their adequacy and to suggest changes to them that might be needed. It was also hoped that the tests, which were performed in the USA and in the UK, would show to regulators, to users of the regulations and to the public that current regulations already provide a very high level of safety for real world accidents. With time, much of the original information regarding these tests and their results has been lost. The few documents that remain have been surveyed and this paper presents summaries from this survey of the tests and their results. [ABSTRACT FROM AUTHOR]

• Performance of full-length UN-SiC fuel rod is modeled during PWR normal operation. • Performance of full-length UN-SiC fuel rod is modeled under PWR LBLOCA and SBO. • SiC irradiation creep is not conducive to the fuel rod hermeticity under LBLOCA and SBO. • Significant post-irradiation annealing of SiC is unfavorable to the fuel rod hermeticity. The adoption of Accident Tolerant Fuels in the US will only likely be realized through improving economics of existing commercial reactors. For this reason, the Nuclear Energy Institute refers to such concepts as Advanced Technology Fuel. This work provides an insight into the performance of the fuel rod with high-density uranium-nitride (UN) pellet in silicon-carbide (SiC) composite cladding for a pressurized water reactor. The UN higher uranium density is one way to overcome the higher cost of SiC relative to current Zirconium based claddings. The material properties and irradiation behavior models of UN and SiC have been incorporated into the BISON fuel performance code. The modeling capability of BISON for SiC cladding is extended with irradiation creep and post-irradiation thermal annealing models. The fuel rod performance during normal operation is evaluated across three fuel cycles with each cycle 24-month long. Selected accident scenarios include a large-break loss-of-coolant accident and a station blackout accident where the potential effects of SiC irradiation creep and annealing rate are emphasized. This work motivates further experimental investigation on irradiation properties of dense UN, SiC irradiation creep and nonisothermal annealing. [ABSTRACT FROM AUTHOR]

• Ba release was observed in Pb-covered sample exposed to air. • Onset of Cs and Ru release was delayed for Pb-covered sample in air. • Cs and Ru were retained in Pb-covered sample exposed to steam. • Pb-covered samples in oxidizing environments have comparatively lower release rates of I. Tests of fission product release under accident conditions were performed on three samples of irradiated UO 2 fuel fragments immersed in molten Pb. Two tests of bare irradiated UO 2 fragment samples were conducted under similar conditions for direct comparison. The irradiated fuel samples had been cooled for several years, then encapsulated and trace-reirradiated to populate limited inventories of short-lived fission products (such as 133Xe and 131I). The tests were conducted with sample temperatures between 1570 and 1590 K. The releases from Pb-covered samples under inert conditions were similar to those from bare UO 2 under inert conditions. Greater 133Xe releases were observed from Pb-covered samples in oxidizing environments. Release of 140Ba was detected for the Pb-covered sample exposed to air. Under oxidizing conditions, Pb-covered samples retained more 137Cs and 103Ru (delayed onset of release in air and complete retention in steam) as well as 131I. [ABSTRACT FROM AUTHOR]

• Under inert conditions, releases of noble gases from ThO 2 were lower than from UO 2. • A sheathed sample of ThO 2 fuel had delayed release of noble gases relative to ThO 2 fragments. • In a steam environment, releases of Kr, Xe, I, and Cs were similar for ThO 2 and UO 2 samples. • No significant releases of Nb, Ru, Ba, or Pa. The accident-conditions fission-product release experiment HCE6 was performed on samples of thoria (ThO 2) fuel and urania (UO 2) fuel irradiated at water reactor operating conditions in the pressurized water loop in the NRU reactor. The tests in the experiment were intended to compare fission-product releases from ThO 2 fuel and UO 2 fuel under similar circumstances, and to explore the behaviour of fission products in ThO 2 fuel under oxidizing conditions. Under inert conditions, or temperatures of ~1720 K, the releases from ThO 2 were significantly lower than from UO 2. A 1920 K test on a sheathed sample of ThO 2 exhibited a delay of ~3000 s in noble gas releases relative to an inert-environment ThO 2 fragment test, possibly due to a decrease in oxygen potential by contact with Zircaloy at high temperature. In a steam environment, the releases of Kr, Xe, I, and Cs, were quite similar for the ThO 2 and UO 2 samples, and the release timing illustrated similar effects of Zircaloy oxidation on the release of these fission products. In an air-steam mixed environment, lower releases of noble gases and volatile fission products from the ThO 2 sample were observed, probably due to the shorter exposure to oxidizing environment and greater amount of Zircaloy present. None of the tests showed significant releases of 95Nb, 103Ru, 140Ba, or 233Pa. [ABSTRACT FROM AUTHOR]