LOCA steam condensation loads in BWR Mark II pressure suppression containment system

Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates ( 30 kg/m²·s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.     

Inerting and venting of mark II containments in BWR nuclear plants

An analysis of hydrogen control systems corroborates containment inerting as the only way of preventing hydrogen explosions which may jeopardize the integrity of BWR Mark II containments during severe accidents. A severe Large Break LOCA and a severe Stuck Open Relief Valve Accident are simulated by the MARCH 2.0 code to compare the advantages and disadvantages of pre-inerting and post-inerting, with or without venting, in BWR Mark II containments.     

The Mark II silicon strip vertex detector and performance of asilicon detector telescope in the Mark II detector at the SLC

A silicon strip vertex detector (SSVD) consisting of 36 independent silicon detector modules has been built for use in the Mark II detector at the SLAC Linear Collider (SLC). The performance of the individual modules and the stability and accuracy of their placement in the mechanical support are discussed. Top gain operational experience at the SLC, a telescope made of three silicon detector modules has been assembled and placed inside the Mark II. Results from the first data run of the SLC on the overall performance of the telescope, including backgrounds, charged particle tracking, and spatial resolution, are presented     

On the failure modes of alternative containment designs following postulated core meltdown

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR representative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best estimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. These include the temperature of the non-condensibles escaping the ice condenser into the upper compartment, the performance of the pressure suppression system, the distribution of non-condensibles between compartments, and the degree and rate of combustion of hydrogen generated from metal-water reactions. For the PWR ice condenser case, results indicate that the containment would be breached by (i) steam overpressurization during the blowdown period (time less than 20 sec) if the ice condenser fails to perform its function, (ii) by overpressurization and thermal stress during the core melt period if 25% or more of the core zirconium reacts with water followed by hydrogen burning and, and (iii) by the overpressurization due to non-condensibles before containment floor penetration is completed. For the BWR Mark III case, similar conclusions can be drawn for the loss of vapor suppression, and for the hydrogen burning if the extent of zirconium-water reaction is more than 35% of the core inventory. If the hydrogen burning fails to materialize, the containment can retain its integrity until containment meltthrough provided the melting is confined to the reactor pedestal area. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded.     

合肥国家同步辐射实验室电子储存环束流损失监测系统中的几个物理问题

电子储存环中损失的束流电子，在撞击真空室壁时，引起“电子-γ-电子”簇射，在撞击点的下游，真空室的外表面形成次级电子（shower电子）的分布。通过探测这些shower电子，可以知道上游某处的束流损失状况。“合肥国家同步辐射实验室加速器二期工程”中的电子储存环束流损失监测系统，就是利用这个原理。在分析了原有加速器束流损失监测系统的缺陷并对国际上各大型加速器进行了调研的基础上，对该系统中的探测器选型、探测器安装位置的选择以及系统的整体结构等物理问题作了阐述。     

Accident scenarios of the TRIGA Mark II reactor in Vienna

The safety report of the TRIGA Mark II reactor in Vienna includes three accident scenarios and their deterministic dose consequences to the environment. The destruction of the cladding of the most activated fuel element, the destruction of all fuel elements and a plane crash were considered scenarios in that report. The calculations were made in 1978 with the software program named STRISK. In this paper, the program package PC Cosyma was applied on the TRIGA Mark II reactor in Vienna and the deterministic consequences of the scenarios to the environment were updated. The fission product inventories of all fuel elements were calculated with ORIGEN2. To get meteorological data of the atmospheric condition around the release area, a weather station was installed. The release parameters were taken from the safety report or were replaced by worst case parameters. This paper focuses on two accident scenarios: the destruction of the cladding of the fuel element with the highest activity content and the case of a large plane crash. The current accident scenarios show good agreement with the calculations from 1978, hence no technical modifications in the safety report of the TRIGA reactor Vienna were necessary. Even in the very worst case scenario – complete destruction of all fuel elements in a large plane crash – the expected doses in the Atominstitut's neighborhood remain moderate.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

A Monte Carlo Simulation of an Actual Segmented Calorimeter: A Study of Calorimeter Performance at High Energies

The computer code system CALOR has been used to simulate data taken with a large segmented liquid argon/iron calorimeter at Fermilab. The resulting energy and angular distributions along with the longitudinal and transverse shower shapes are compared to experimental data in the range 1-38 GeV. Results are presented extending the Monte Carlo simulations to 125 GeV. The energy resolution of an incident hadron can be determined with a resolution ?E/E = (7.6 + 29.2/?E)% and its direction with a resolution of ?(?) = 22.7 + 390/E) mrad (where E is in GeV).     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

Ignitors to mitigate the risk of hydrogen explosions — A critical review

Risk analyses and the accident at the Three Mile Island plant (TMI-2) have shown that the formation of large amounts of hydrogen during severe accidents poses a real danger to the containment integrity. Proper means must be taken to prevent the occurance of global or sometimes even local detonations. The applicability of ignitor systems to protect large dry PWR containments is critically discussed. Although already adopted for Mark III-BWR and Ice-condenser PWR-containments the reliability of spark plug ignitor systems to protect large dry containments from the possible consequences of a local or global detonation is neither proven experimentally nor analytically. An experimental study of possible post-accident inertisation procedures deems necessary and may yield a more convincing mitigation procedure.     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

Analytically predicted versus measured response of a free-standing steel containment vessel subjected to safety-relief valve discharge loads

Following the actuation of safety-relief valves in BWR nuclear power plants, first water then air and steam are cleared from the discharge lines through quencher devices into a suppression pool. This clearing results in water spike, air bubble, and condensation pressure loads applied to structures in the pool, and the surrounding containment vessel.The Leibstadt Nuclear Power Plant has the only free-standing steel Mark III containment vessel in the world. All other steel Mark III containment vessels have concrete backing in the suppression pool region, which dampens clearing load responses. As such, it is of interest to note how this steel vessel responds to discharge pressures, and compare these responses to analytically predicted results.The purpose of this paper is to compare the analytical results used to design the steel containment vessel with the responses measured during in-plant testing. The analytical methods considered the effects of fluid-structure interaction. The test program included initial and consecutive actuations of a single valve, and initial actuation of multiple (four) valves. The conclusion of the comparison is that, in general, there are large conservatisms in the analytical predictions versus measured responses.     

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.     

THIBO experiments — Thermohydraulically induced fuel pin oscillations in Na-cooled reactors

Unusual wear marks and damage to pins and spacer grids have been found in KNK II fuel element. For instance, there was a pronounced interaction not only between the spacers and the cladding, but also among the fuel pins outside the spacer grids. There is a theory postulating that the causes are not hydraulically induced vibrations, but powerful low-frequency pin oscillations created by special thermohydraulic and geometric conditions. This implies that the pin power, the sodium mass flow and velocity, the pin clearance within the spacer grid system, and the pin structure play the decisive roles.This phenomenon is so interesting that we are performing out-of-pile experiments in order to contribute to the verification of the theory and to learn something about the mechanism and the limiting conditions.The test section simulates a single fuel pin in a KNK II subassembly. There is an electric heater with dimensions of Mark II type pins. All spacers can be adjusted very precisely by micrometer screws. A large number of thermocouples indicate the sodium temperatures around the heater. An X-ray system allows part of the heater to be made visible.In 1988, we succeeded in making the fuel pin simulator oscillate with different parameters. The accompanying azimuthal temperature oscillations grew to more than 100 K and a period of more than 10 s. Maximum pin bending, as determined by X-raying, was nearly one millimeter. The reproducibility of the oscillations was remarkably good.     

Is Mark I shell failure really important? — Part one

This two part paper discusses Mark I shell failure from two perspectives. In Part One a general overview of nuclear health consequences and risks is provided. Comparisons between calculated risks and the NRC's safety goals are offered. Use is made of present PRA results to calculate risks as well as simpler techniques that do not rely on more realistic calculations of source terms. These simpler techniques for estimating nuclear risks also minimize the use of PRA methodologies.In Part Two the results of specific Mark I accident sequences is analyzed. Source terms, accident containment pressure transient plots, and a discussion of the relevance of this information to the Mark I shell failure is provided. Of particular interest is the mitigation potential of extended use of the automatic depressurization system during station blackout conditions.     

远距离大动态范围电荷测量电路的设计

介绍了大型高海拔空气簇射观测站(Large High Altitude Air Shower Observatory,LHAASO)空气簇射芯探测器阵列(Shower core detector array,SCDA)读出电子学方案的预研设计。系统采用基于电荷积分法的电荷测量方案,读出电子学通过同轴电缆接收光电倍增管输出的电流信号;采用在输入端与电荷积分放大器的虚地点之间接入等效50?电阻的终端阻抗匹配方案,并通过Pspice仿真验证该阻抗匹配的可行性。电路测试结果表明,该电路能满足远距离10 bit大动态范围电荷测量的设计指标要求。     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Performance of a Thallium Chloride Scintiliator as a Total Absorption Shower Cascade Detector

Studies have been made of the performance of a large sample of the new heavy scintillator TlCl(Be,I) as a total absorption shower cascade (TASC) detector for high energy electrons and gamma rays. The observed energy resolution for electrons in the range 0.1 to 15 GeV is reported and compared to a Monte Carlo calculation of expected fluctuations in energy leakage from the crystal volume. The performance of this new scintillator and its potential applications are discussed.     

A zero dimensional model for simulation of TRIGA Mark II dynamic response

Aim of this work is to reproduce the dynamic behavior of the TRIGA Mark II reactor of the University of Pavia on the entire operative power range (i.e. 0–250 kW) using a zero dimensional approach. In this work the coupling between neutronics and thermal-hydraulics in natural circulation has been considered. In specific, a point reactor kinetics model with one energy group and six delayed neutron precursors groups has been adopted while for thermal-hydraulics modeling two regions have been defined (i.e. the fuel and the coolant). The nonlinear system of coupled Ordinary Differential Equations has been solved by means of MATLAB Simulink^®, which represents a reliable tool for dynamic and control analysis. The model has then been validated through the comparison with a set of experimental data collected in four different reactor power transients, obtaining a very satisfying agreement. Finally, the linear stability analysis of the TRIGA reactor has been performed by means of the root locus, finding out that the power level at which reactor is operating deeply influences the position of the poles of the transfer function between control rod height and neutron density. These results can then be employed as a reliable starting point in designing an automatic device for reactor power control.     

Finite element dynamic bifurcation buckling analysis of torispherical head of BWR containment vessel subjected to internal pressure

In this paper the bifurcation buckling pressure for the torispherical head of the Mark II type BWR containment vessel subjected to dynamically applied internal pressure is calculated, using a finite element program for a dynamic analysis. Three kinds of dynamic loadings, that is, step loading, ramp loading and pulse loading are considered in the present analysis. The minimum bifurcation buckling pressure is predicted for the respective loadings. The minimum bifurcation buckling pressure for dynamic loading is much lower than the bifurcation buckling pressure for static loading.