Leak-before-break assessment of CANDU pressure tube considering leak detection capability

The leak before break (LBB) concept is employed as defense in depth in CANDU reactors in order to avoid an unstable failure in pressure tubes. The LBB in pressure tubes can be ensured when detection, confirmation and location of the leak are carried out and the reactor is placed in a depressurized condition before the crack exceeds critical crack length. Leak detection and location is provided by the annulus gas system (AGS). Therefore, the evaluation parameters for LBB assessment, such as leak detection and leak location capabilities, should be made available through AGS performance test. Recently, the AGS in-situ tests with a simulated moisture injection were carried out in one of the CANDU reactors in Korea. This paper presents the LBB assessment performed taking into account the leak detection capability and the modification of operating procedure therefrom.     

基于GEM工艺的裂变时间投影室性能模拟研究

基于GEM工艺的裂变时间投影室具有探测效率高和空间分辨率高的特点,可实现裂变产物的多参量测量。本文主要研究基于GEM工艺的裂变时间投影室在不同条件下的测量精度,使用Garfield++软件计算得到裂变时间投影室中不同的裂变产物质量数测量误差约为4～6 u,并通过时间信息的径迹重建研究了裂变碎片在不同工作气体中的角度分辨。研究发现,电子漂移时间长的工作气体中,裂变产物具有更好的角度分辨,并可依此在实验中选择合适的工作气体、气压和漂移电场强度来进行裂变碎片的测量。     

液态燃料氯盐快堆最小需求燃耗分析

为探究采用增殖燃烧模式运行的液态燃料氯盐快堆的平均卸料燃耗深度，基于中子平衡分析方法，选取5种常用氯盐，提出在线清除裂变气体和难溶裂变产物方案来维持增殖燃烧运行模式，主要研究分析了氯盐的重金属密度和在线处理方案对最小需求燃耗的影响以及无限栅元模型下维持增殖燃烧模式可接受的堆芯中子损失项。分析表明68NaCl-32UCl₃和20UCl₃-80UCl₄的最小需求燃耗分别是30.47%FIMA（FIMA是指已裂变原子数与初始的总装料金属原子数之比）和10.28%FIMA；清除裂变气体和难溶裂变产物后，60NaCl-40UCl₃可接受的中子损失项从3.49%提高到10.68%。结果表明氯盐的重金属密度对最小需求燃耗有明显影响，同时清除裂变气体和难溶裂变产物能够较大提高燃料盐系统的中子经济性，以及提高增殖燃烧模式运行可接受的堆芯中子损失项。      

Development of fuel failure detection system for a high temperaturegas cooled reactor. II

For pt.I see ibid., vol.34, p.567 (1987). A fuel failure detection (FFD) method based on selective detection of short-life and gaseous fission products, developed for a high-temperature gas-cooled reactor, is described. An improved precipitator was used as a detector for the fission products and the performance of the FFD system was tested using an irradiation rig at the Japan Material Testing Reactor. In the rig, three kinds of samples of coated-particle fuels were irradiated and each sample of the primary helium gas was fed to the FFD system. Failure rates of the three fuel samples called intact, normal, and slightly failed, were estimated at about 10^-6, 10^-5, and 10 ^-4, respectively. The FFD system showed a significantly increased response in counting rate for the sample gas with a failure rate of 10^-4. The FFD system did not respond to the sample gas with the smaller failure rate of 10^-5 even when the background level of long-life fission products in the primary coolant gas increased with fuel temperature and reactor power     

Extending FEAST-METAL for analysis of low content minor actinide bearing and zirconium rich metallic fuels for sodium fast reactors

Computational models in FEAST-METAL fuel behaviour code have been upgraded to simulate minor actinide bearing zirconium rich metallic fuels for use in sodium fast reactors. Increasing the zirconium content to 20-40 wt.% causes significant changes in fuel slug microstructure affecting thermal, mechanical, chemical, and fission gas behaviour. Inclusion of zirconium rich phase reduces the fission gas swelling rate significantly in early irradiation. Above the threshold fission gas swelling, formation of micro-cracks, and open pores increase material compliancy enhance diffusivity, leading to rapid fuel gas swelling, interconnected porosity development and release of the fission gases and helium. Production and release of helium was modelled empirically as a function of americium content and fission gas production, consistent with previous Idaho National Laboratory studies. Predicted fuel constituent redistribution is much smaller compared to typical U-Pu-10Zr fuel operated at EBR-II. Material properties such as fuel thermal conductivity, modulus of elasticity, and thermal expansion coefficient have been approximated using the available database. Creep rate and fission gas diffusivity of high zirconium fuel is lowered by an order of magnitude with respect to the reference low zirconium fuel based on limited database and in order to match experimental observations. The new code is benchmarked against the AFC-1F fuel assembly post irradiation examination results. Satisfactory match was obtained for fission gas release and swelling behaviour. Finally, the study considers a comparison of fuel behaviour between high zirconium content minor actinide bearing fuel and typical U-15Pu-6Zr fuel pins with 75% smear density. The new fuel has much higher fissile content, allowing for operating at lower neutron flux level compared to fuel with lower fissile density. This feature allows the designer to reach a much higher burnup before reaching the cladding dose limit. On the other hand, in order to accommodate solid fission product swelling and to control fuel clad mechanical interaction of the stiffer fuel, the fuel smear density is reduced to 70%. In addition, plenum height is increased to accommodate for fission gases.     

基于弥散燃料颗粒开裂的裂变气体释放模型

根据弥散燃料颗粒开裂后裂变气体的3种释放途径,分别建立了裂纹连通释放模型、气泡连通释放模型以及原子扩散释放模型,综合得到了基于弥散燃料颗粒开裂的裂变气体释放模型,并采用该模型对裂变气体释放量进行了计算。结果表明:裂变气体释放量主要由裂纹连通释放途径贡献;燃耗深度越高,裂变气体释放量的增加速率会越大;随着退火温度的增加,裂变气体释放量迅速增加,而退火时间越长,裂变气体释放量的增加速率越低。通过裂变气体释放量模型计算得到的裂纹宽度与实验观察到的裂纹宽度符合较好,对比结果验证了基于弥散燃料颗粒开裂的裂变气体释放模型的合理性。      

An evaluation of pin-to-pin failure propagation due to fission gas release in fuel subassemblies of liquid-metal-cooled fast breeder reactors

A survey is given of work performed to evaluate the potential for pin-to-pin failure propagation due to fission gas release in fuel subassemblies of LMFBRs. Reference is made to publications available in the open literature; recent experiments and analyses are dealt with in more detail.Two distinct failure propagation mechanisms are identified, namely: (1) thermal transients, and (2) mechanical loads. It is concluded that rapid and extensive pin-to-pin failure propagation due to fission gas release is not possible, even for the very high gas release rates considered.     

Thermodynamic consequences of leakages in a double containment during severe accidents

For a large nuclear power plant under normal operating conditions a leakage rate for the containment of 0.25 vol.%/day is admissible. During a successfully controlled LOCA leakages of the containment will be released through filters by the annulus^* air exhausting system into the environment. During a core melt accident a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. When openings in the containment steel shell occur before a catastrophic failure, a depressurization into the annulus takes place. The area of the openings determines the depressurization rate and the thermodynamic conditions in the annulus. Furthermore the behaviour of the components being necessary for accident mitigation is influenced too. This paper discusses the thermodynamic consequences of leaks in the containment shell of a German PWR during a core melt accident. The results of those calculations are the necessary boundary condition for the estimation of fission product retention in the annulus.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Radiation re-solution of fission gas in uranium dioxide and carbide

Calculations have been performed to estimate the removal rate of fission gas atoms from bubbles due to collisions with energetic fission fragments and recoil cascades. The efficiency of this process was found to be higher than estimated earlier, but is still too low to be responsible for the experimental observations of fission gas bubble destruction during irradiation of oxide fuel. An irradiation experiment to investigate the interaction of fission spikes with free surfaces has enabled a simple theory to be developed which can explain the shrinkage of bubbles and pores by the surface relaxation of a shock wave produced by the passage of a fission fragment. This mechanism occurs in oxides but not carbides because of the faster dispersion of the fission fragment energy and provides the major reason for the difference in gas bubble distributions in oxide and carbide fuel. This process, however, does not remove gas atoms from the bubbles. Since high levels of apparently diffusive fission gas release are observed in oxides, the “effective solubility” of the fission gases required for this release must be sought in phenomena other than the fission spike.     

Behaviour of fission gas products in single crystal UO2 during intermediate temperature irradiation

A new mathematical interpretation is presented of fission gas release from monocrystalline uranium dioxide fuel during intermediate temperature irradiation in terms of a defect trap model, knock-out process and diffusion of bubbles. In the present model it is assumed that gas in an intermediate state exists side by side with the dissolved fission gas and that trapped in bubbles. It is assumed also that the isolated gas atoms, being re-dissolved, are immobile.The present model gives a satisfactory interpretation of the relative proportions of isotopes in the steady state fission gas release at diffrent temperatures. The dependence of fractional fission gas release on fission rate is also interpreted; regimes either proportional to fission rate or inversely proportional to fission rate are predicted depending on the fission rate interval considered. Both temperature dependent and temperature independent fission gas release can arise.The presented dynamic method of studying the release of fission gases during irradiadion provides a further test beside the static method of the veracity of the assumed mechanisms. Calculations show that fission gas behaviour becomes more complex for oscillated fission rate in the regime where the fractional release is inversely proportional to the fission rate for the steady state.     

Irradiation testing of design models for the GCFR fuel pressure equalization (vent) system

The major consideration in the design of the pressure equalization system for the gas-cooled fast breeder reactor is the release and venting of gaseous and volatile fission products. Single vented rods have been irradiated in the thermal flux of the Oak Ridge Research Reactor (ORR) at GCFR operating conditions of 12–15 kW/ft and 565–685°C cladding outside temperature to determine the fission product release and to verify the design concept. Results obtained to date from measurements of fission gas release and transport have been compared with predictions based on design assumptions to verify analytical models and have established a degree of conservatism of design assumptions.The release of radioactive gases from the fuel matrix was measured directly at 12 kW/ft in an operating fuel rod and was found to be significantly less than the design assumption of 100% instantaneous release and less than predictions using the diffusion model with Findlay's coefficients. Although solid state diffusion was found to be the dominant process delaying the venting of fission gases in the experimental irradiation, fission gas interdiffusion in helium will be the dominant venting transport process for the reactor design. Delay of fission gases by adsorption on charcoal was verified at trap operating temperatures for burn-ups up to 54 000 MWd/t. Volatile fission products (cesium and iodine) did not migrate beyond the fuel-blanket interface. The feasibility of the vented-fuel-rod design concept has been established.     

Investigation of the impurities and fission products in the AVR coolant gas at an average hot-gas temperature of 950°C

Since 27 February, 1974, the AVR pebble bed reactor has been producing gas at an average temperature of 950°C. Therefore it is possible for the first time to gain experience in high temperature reactor operations and experiments with such a high temperature level. This is of particular interest with regard to efforts using high temperature reactors for production of nuclear process heat. This paper reports briefly on the preparations for a temperature increase and on the first experimental results obtained with a hot-gas temperature of 950°C. Measured data are given on the behaviour of inactive gaseous impurities, on the increase of fission gas activities, and on the increase of concentrations of solid fission products in the helium coolant gas. While the activities of the fission gases showed an insignificant increase in the coolant gas, considerable increase of activity was measured for solid fission products, especially for Ag isotopes. However, activities released from fuel elements are low so that there are no operational or safety problems.     

Validation of simplified fracture mechanics methods by testing of real components

The evaluation of integrity of structural components is often based on the proof of leak-before-break (LBB). Leak-before-break behaviour in piping constitutes a fail-safe condition. Which means that, during multiplied loading conditions, a defect results at first in a leakage. The crack length which leads to the leakage is smaller than the critical through-wall crack length. Simplified fracture mechanics concepts are used for the demonstration of LBB. For this the conservative, safe calculation of the critical through-wall crack length for ductile failure is necessary. To validate simplified calculation methods for circumferential cracks (flow stress concept (FSC); plastic limit load (PLL)) and for axial cracks (Battelle approach (BMI); Ruiz approach (RUIZ)) all available experiments on real structural components, especially on pipes, were analysed and evaluated by the mentioned simplified methods (approximately 460 experiments). The methods were adapted by application of correction factors, mainly on the flow stress, to result in conservative (safe) and realistic (as near as possible to the experiments) predictions. Depending on method (FSC, PLL, BMI, RUIZ), crack orientation (circumferential and axial cracks) and type of material (ferritic and austenitic material) different definitions of flow stresses were established.     

The measurement of the radioactivity of short-lived gases in a mixture of fission products

The article describes a procedure for measuring the activity of short-lived isotopes of inert gases in a mixture of fission products of U²³⁵. The daughter products of the gases are selectively accumulated in filters set up in a flow-through system and are measured by a radiometric unit. A sensitivity of the order of 10^–15 grams of inert gas is obtained. The procedure is intended for the investigation of the discharge of gaseous fission fragments from fuel elements.The author takes this opportunity to express his gratitude to V. M. Makurin and S. N. Stepanov for their participation in the conduction of the experiments.     

Artificial neural networks in CT–PT contact detection in a PHWR

In a pressurized heavy water reactor (PHWR), contact between calandria tubes (CT) and pressure tubes (PT) makes them susceptible to delayed hydrogen cracking. Periodic inspection of the channels must be carried out to detect the contact. As the number of channels in a PHWR is very large (306 in a 230 MW plant) periodic in-service inspection of all the channels leads to an unacceptable downtime. A non-intrusive technique that employs a system identification method is presently used for contact detection. The technique tends to overpredict the number of channels in contact, i.e. they diagnose many channels as contacting while the channels are in fact not in contact. This puts a large number of healthy channels on the at risk list reducing the efficacy of the method. This paper demonstrates the power of artificial neural networks (ANNs) in diagnosing the CT–PT contact. A counterpropagation neural network consisting of a Kohonen layer and a Grossberg layer has been employed. The noise tolerance of the network has been demonstrated.     

Studies of Ceramic Fuels with Use of Fission Gas Release Loop, (IV)

In-pile release of fission gas from sintered UC pellets in the presence of 8–230 ppm of water vapor in the He sweep gas was measured over the temperature range of 160°–1,000°C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established that the release of short-lived fission gases during irradiation was controlled mainly by pseudo-recoil, while chemical reaction between UC and water vapor, as well as knock-out, appeared to contribute much more significantly in the case of the longer-lived fission gases. The release of fission gas after reactor shutdown was shown to be governed by the UC-H₂O reaction. The ratio of the release due to this reaction in reference to the total release was found to be dependent not only on the concentration of the water vapor but also on the amount present of the accumulated reaction products. Also, a discussion is given on the inordinately high release of ^135mXe observed at 600°C.     

Heavy-gas injection in the Generation IV gas-cooled fast reactor for improved decay heat removal under depressurized conditions

Decay heat removal is a key safety and design issue for the Generation IV gas (helium)-cooled fast reactor. This paper investigates the natural convection capability of the dedicated DHR loops under depressurized conditions while injecting a heavy gas into the system. Investigated is a loss-of-coolant accident using the TRACE code. The goal of the study is to improve fuel/cladding temperature behavior during LOCA transients with the enhancement of passive safety by operation in natural convection only, while accepting 10 bar back-up pressure in the guard containment. The paper investigates the cooling capabilities of different heavy gases. Furthermore, different injection locations and mass flow rates have been tested, in order to address possible core-overcooling problems resulting from rapid depressurization of the gas reservoir. It has been shown that, among the gases investigated, CO₂ is the best choice from the thermal-hydraulics viewpoint, being able to cool the core satisfactorily for a broad range of injection rates. N₂ can be envisaged as an alternative solution in case of chemical problems with CO₂. Supplementary studies carried out for the CO₂ and N₂ injection cases include that of the sensitivity to the number of available DHR loops and to the LOCA break-size. The effect of the resulting neutron spectrum changes on the shutdown-reactivity margin has also been investigated.     

Experimental and theoretical investigations of KWU on leak-before-break behaviour and its application on excluding of breaks in nuclear piping of KWU-plants

The fracture of pipes with longitudinal and circumferential cracks was investigated by experiments and theoretical approaches (flow stress criteria and limit load analyses).The experiments show that the critical crack dimensions can conservatively be determined by fracture mechanics.The tests and calculations are applied to the primary coolant piping with hypothetical longitudinal and circumferential defects. Reactor systems, design, fabrication, stress analysis, material, non-destructive testing, quality control and inservice inspection are considered referring to the leak-before-break behaviour. On the basis of the extreme toughness of the materials, the known loads, the high level of non-destructive examinations, the leakage monitoring system and the high quality of manufacture and processing it is shown that a spontaneous failure need not to be postulated.