Assessment of integrity of components in piping of 500 MWe PHWR: using R-6 method

The primary heat transport system of 500 MWe Indian PHWR comprises of straight pipes, elbows and headers. A study was conducted to qualify piping system for leak before break. R-6 method was used to assess integrity of system for leakage size crack (LSC), the margins on crack initiation load and unstable crack growth loads. Option 2 (material specific failure assessment diagram), Category 3 (ductile tearing) analysis was used for straight pipes, elbows and header. In order to enhance the confidence in the analytical results, detailed sensitivity analysis was also performed. For sensitivity analysis, variation in material properties, LSC was considered. The effect of variation in temperature on material properties was also considered. Tensile and fracture properties used for base and weld material data were generated from pipe material obtained from 220 MWe Nuclear Power Plant, under construction.     

Detecting small leaks in pipes in nuclear power plants by measuring aerosol parameters

It is found experimentally that when through defects appear in water pipes the concentration and dispersion composition of the aerosols in the room air change. A water or steam leak in a pipe at an early stage of the development of a defect, when the leak still has no effect on the standard operation of the equipment, can be detected by monitoring the aerosol component of the air. The stage where a leak can be detected by aerosol monitoring depends on the background aerosol concentration in the room air, the temperature and pressure in the pipe, and the arrangement of samplers in the monitoring system. The leak-detection method proposed is much more sensitive than moisture content measurements.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 189–195, September, 2004.     

Acoustic leak detection for reactor coolant systems

An acoustic leak detection (ALD) facility has been constructed at Argonne National Laboratory (ANL), and tests have begun on five laboratory-grown cracks (three fatigue cracks and two thermal-fatigue cracks) and two field-induced intergranular stress-corrosion cracks (IGSCC). The acoustic characteristics of the IGSCC were established and the minimum leak rate that is detectable at a distance of 50 cm under laboratory conditions was identified to be 0.001 gal/min for a frequency window of 300–400 kHz. Cross-correlation techniques were also demonstrated for locating leaks and distinguishing among leak types. A comparison of acoustic spectra of the two leaking IGSCC at flow rates of 0.005 gal/min revealed nearly identical signals in the frequency range from 200 to 400 kHz. Below 200 kHz, the differences in geometry between the two cracks can be reflected in a signal difference. These data suggest that geometrical effects may be less significant at frequencies above 200 kHz. It appears that leak rates may be more reliably related to acoustic signals at higher frequencies.     

大型加压重水反应堆隐蔽攻击方法研究

为了对大型加压重水反应堆（PHWR）安全防御系统研发提供帮助,研究了PHWR网络控制系统中潜在的攻击方式,并提出了一种基于樽海鞘群优化高斯过程回归算法的隐蔽攻击方法。该方法在对PHWR网络控制系统实施虚假数据注入时,通过樽海鞘群优化高斯过程回归算法进行系统辨识,获得PHWR受攻击区域高精度的估计模型,并利用该估计模型实现隐蔽攻击。仿真结果表明,该攻击方法对PHWR造成一定破坏性的同时具有高度的隐蔽性能。      

Thermal stress analysis of containment building in case of a Main Steam Line Break (MSLB)

In Indian PHWR, containment building is one of the primary barriers for mitigating the consequences of a Loss of Coolant Accident (LOCA), and Main Steam Line Break (MSLB) accident. It is desired to know the temperature transients as well as the resulting thermal stresses in the containment structures of 220MW(e) PHWR, Kaiga Nuclear Power plant under postulated MSLB event. The high enthalpy steam discharged into the containment space comes in contact with the Structural Wall (SW) of containment, Inner Containment Wall (ICW) and Raft. The containment wall temperature rises due to heat transfer from steam-air mixture. To calculate the transient temperature distribution across the containment walls, it is necessary to determine containment ambient temperature and heat transfer coefficient for the condensing steam on the internal structures. Hence, at the outset, a thermal hydraulic code was developed to predict the pressure-temperature transients and condensation heat transfer coefficient transients (using various condensation models) based on the mass and energy of high enthalpy steam released into containment. The effect of various condensation models on containment pressure-temperature was evaluated. The thermal boundary conditions such as containment temperature and heat transfer coefficient, evaluated from the thermal hydraulic code using Uchida condensation model, were subsequently applied as boundary conditions to a two-dimensional axi-symmetric containment model developed using a FEM code for estimating two-dimensional temperature profiles and the resulting thermal stresses.     

Analysis of PHWR lattice experiments with organic coolant by the computer code club

An efficient computer code club, based on the combination of a small-scale collision probability and a large-scale interface current method, was developed for the analysis of pressurized heavy-water reactor (PHWR) lattice cells. A large number of experiments with different fuel clusters and D₂O and air coolants were analysed using this code. The results were found to be very encouraging. However, when club was used for analysing experiments with organic coolants, the results were found not to be in good agreement with the experiments. This paper discusses the reasons for this and proposes a remedy. Finally, it gives the results of the analysis of these experiments with the modified computer code club.     

Accident progression within the secondary containment boundary of large LMFBRs

This study deals with the sodium spillage phenomenon as it relates to accident energetics and containment integrity. Sodium spillage has been identified as an important issue for large LMFBRs because of the large inventory of sodium present and the potential for energetic accidents. Energetic core-disruptive events leading to slug impact could open leak paths in the reactor cover and vent sodium into the secondary containment. Sodium fires in the containment building could lead to pressurization and thermal stressing of the surrounding structure and jeopardize containment integrity. The potential consequences of such a scenario have prompted the development of analytical tools to quantify the spillage process.One of the primary concerns in assessing the integrity of secondary containment is the amount and velocity of sodium which may be ejected from the primary vessel. A parametric study has been performed, the purpose of which was to study the sensitivity of sodium spillage to accident energetics. Treatment of the spillage process was accomplished with the ICECO code employing a quasi-Eulerian method. A 1000 MWe reactor, with prescribed leak paths, is modelled and analyzed during the slug impact phase. Leak paths are assumed to exist as annular penetrations in the reactor cover and as a gap at the vessel-head junction. The behavior of sodium spillage is investigated under conditions of different accident energetics, various opening sizes, and multiple leak paths, with both stationary and moving reactor covers. The relative influence of short and long term spillage is also addressed.During the transient period immediately following slug impact it was found that spillage from annular penetrations in the reactor cover is only weakly sensitive to changes in slug velocity. The same conclusion applies to spillage from a fixed gap at the vessel-head junction. Significant sensitivity of spillage to accident energetics was seen only in cases of spillage from the vessel-head junction when the reactor cover was movable. The influence of slug impact on the motion of the reactor cover leads to the conclusion that sodium spillage is most sensitive to accident energetics inasmuch as it affects the size of the leak path.     

Optimization of depleted uranium bundle loading in fresh core of Indian PHWR by evolutionary algorithm

This paper is concerned with the Indian design of a 220 MWe pressurized heavy water reactor (PHWR) having natural uranium (NU) fuel and heavy water as moderator and coolant. At the beginning of life, it is necessary to flatten the power by loading some depleted uranium (DU) bundles to achieve a nearly full power operation. The determination of best possible locations of DU bundles, which maximize fuel economy as well as safety, is a large-sized combinatorial optimization problem with constraints. In the past, 384 DU bundles have been loaded in locations determined by manual intuition in an Indian PHWR and maximum permissible power of 93% full power (FP) was obtained. In the present paper, a modern evolutionary algorithm called estimation of distribution algorithm (EDA) is used to improve upon this distribution. Optimum distributions of DU bundles which maximize K_eff and give 100% FP without violating safety parameters such as maximum permissible bundle power, channel power, channel outlet temperature and permitted reactivity worths of shut-down systems are obtained. Another aspect studied in this paper is to find out how far one can increase the number of DU bundles loaded in the core. This will minimize the NU bundles requirement, extract more power from DU bundles and thus provide better fuel utilization. The idea is to conserve NU bundles. The optimum distribution of DU bundles has been obtained for the total number of DU bundles ranging from a few hundreds to a few thousands. It is found that, depending on various conditions, about 60–80% of the core can be loaded with DU bundles leading to a substantial saving in NU bundles. Some variation in the implementation of EDA to generate loading pattern of PHWR reactor core is also studied.     

Development of Acoustic Leak Detection and Localization Methods for Inlet Piping of Fugen Nuclear Power Plant

The development work carried out on Fugen NPP is focused on detection of a small leakage on the reactor's inlet feeder pipes at an early stage by an acoustic leak detection method with usage of high-temperature resistant microphones. Specifically, the leak rate of 0.046 m³/h has been chosen as a target detection capability for this system. A cross-correlation technique has been studied for leak detection under low signal-noise ratios. The study shows that the sound diffusion on piping causes distortion of leak signals that results in their low correlation. A leak-location estimator and multi-channel correlation value, associated with estimated leak position, have been employed to detect such low-correlated leak signals. A method based on cross-correlation of signal spectral components has been proposed to deal with non-stationary leak signals. Joint-Time-Frequency-Analysis has been applied to analyze such signals, whilst a Wavelet decomposition technique has been used to extract their short-term spectral fluctuations. Since the spectral components are less affected by signal distortion, they provide higher correlation value and can be applied for leak detection under lower signal-noise ratios. The possibility of detecting and locating a small leakage by the methods proposed has been demonstrated by a number of simulation tests conducted on the Fugen NPP site.     

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.     

Adequacy of Power-to-volume Scaling Philosophy to Simulate Natural Circulation in Integral Test Facilities

Theoretical and experimental investigations were carried out to study the adequacy of power-to-volume scaling philosophy for the simulation of natural circulation and to establish the scaling philosophy applicable for the design of the Integral Test Facility (ITF-AHWR) for the Indian Advanced Heavy Water Reactor (AHWR). The results indicate that a reduction in the flow channel diameter of the scaled facility as required by the power-to-volume scaling philosophy may affect the simulation of natural circulation behaviour of the prototype plants. This is caused by the distortions due to the inability to simulate the frictional resistance of the scaled facility. Hence, it is recommended that the flow channel diameter of the scaled facility should be as close as possible to the prototype. This was verified by comparing the natural circulation behaviour of a prototype 220MWe Indian PHWR and its scaled facility (FISBE-1) designed based on power-to-volume scaling philosophy. It is suggested from examinations using a mathematical model and a computer code that the FISBE-1 simulates the steady state and the general trend of transient natural circulation behaviour of the prototype reactor adequately. Finally the proposed scaling method was applied for the design of the ITF-AHWR.     

Development of the detecting method of helium gas leak from canister

In the concrete cask storage system, spent fuel is installed and weld-sealed in a cylindrical container called a canister. The canister is filled with helium gas and its containment shall be maintained and inspected during storage. The helium gas enhances heat removal from spent fuel. When the helium gas leaks, the effect of helium gas convection is weakened in the canister. Thereof, the temperature on the canister surface changes.In present tests, it was found that temperatures of the center of the top and the bottom on the canister surface change remarkably during the helium gas leak. Therefore, we defined the temperature difference as ΔT_BT. And one can detect helium gas leak using the change of ΔT_BT. ΔT_BT increases monotonously toward a constant value during helium gas leak, even if the inlet air temperature drops. The helium gas leak can be detected at the early stage of the leak by observing both ΔT_BT and inlet air temperature.     

Study on Risk-Informed In-Service Inspection for BWR Piping

Risk-informed in-service inspection for piping was studied for a BWR plant. Piping segment rupture probabilities were determined by Bayesian transform from piping failure events in the database of the OECD-NEA Piping Failure Data Exchange project. Based on the methodology of the Westinghouse Owners Group, core damage frequency induced by each segment rupture was determined by the use of a surrogate component in the PSA model. Nondestructive examinations were added to leak examinations for segments of the resultant high risk significance. The changes from current examinations gave around 29% reduction of segments subject to both leak and nondestructive examinations within the total segments. Deterministic insights and engineering judgments on top of risk significance should be applied to obtain the final decision of inspection methods. An extent-of-examination was studied by the adoption of the Perdue-Abramson model in the Westinghouse Owners Group methodology. The necessary leak frequency of a crack in a segment was calculated by the probabilistic fracture mechanics code PRAISE. Two segments of high risk significance showed lower or slightly higher extents-of-examination, respectively, than the current extent-of-examination. To contribute to the enhancement of the scientific rationality of piping inspections, technical knowledge was accumulated.     

重水堆3H排放量控制值的建议

为了满足急需,根据重水堆核电站的实际释放资料,参照国际GB6249-86对压水堆流出物中放射性核素释放的控制水平,并考虑到干什么山厂址多堆型、多机组的特点,对可适用于秦山三期重水堆核电站的3H排放量控制值提出了建议,并对通用控制值可能的取值范围进行了讨论.     

Evaluation of methods for leak detection in reactor primary systems

No currently available, single leak-detection method combines optimal leakage detection sensitivity, leak-locating ability, and leakage measurement accuracy. Technology is available to improve leak detection capability at specific sites by use of acoustic monitoring. However, current acoustic monitoring techniques provide no source discrimination (e.g., to distinguish between leaks from pipe cracks and valves) and no leak-rate information (a small leak may saturate the system).Seven cracks, including three field-induced IGSCC specimens and two thermal-fatique cracks, have been installed in a laboratory acoustic leak detection facility. The IGSCC specimens produce stronger acoustic signals than the thermal-fatigue cracks at equivalent leak rates. Despite significant differences in crack geometry, the acoustic signals from the three IGSCC specimens, tested at the same leak rate, are virtually identical in the frequency range from 300 to 400 kHz. Thus, the quantitative correlations between the acoustic signals and leak rate in the 300–400 kHz band are very similar for the IGSCC specimens. Also, acoustic background data have been acquired during a hot functional test at the Watts Bar PWR. With these data, it is now possible to estimate the sensitivity of acoustic leak detection techniques.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.     

Development of water leak detection method in fusion reactors using water-soluble gas

A water leak detection method using krypton (Kr) as a water-soluble tracer has been proposed for fusion reactors with fully circulating the in-vessel cooling water. This method was targeted for applying to the International Thermonuclear Experimental Reactor (ITER), and the 10⁻³ Pa m³/s order of water leak valves were fabricated and connected to the water loop circuit. The water leaks were effused into the vacuum vessel evacuated by a cryopump and the water dissolved Kr was detected by a quadrupole mass spectrometer (QMS). When two leak valves with 1 m distance were attached to the test pipe with 30 °C heating, two distinct, the mass to charge number ratio (m/e) of 84 peak current rises caused by the water leak were successfully detected with the time interval of 39 s. On the other hand, the water accession length as a function of the traveling time was calculated by considering a natural convection flow caused by the 30 °C heating, where the traveling time was 44.6 s for the 1 m length. This means that the observed positional accuracy is 12.6%, based on the calculation. To enhance the positional accuracy, the detailed flow simulation is indispensable. This method can be applied to the ITER condition.     

秦山重水堆核电厂的主要设计改进

为了改善电厂性能 ,在CANDU 6设计的基础上作了 90余项设计变更和改进 ,使其在设计上已成为目前世界上在建造、运行的CANDU 6机组中最好的重水堆核电厂。这些设计改进对同类核电厂具有重要的参考价值     

Flexible path optimization for the Cask and Plug Remote Handling System in ITER

The Cask and Plug Remote Handling System (CPRHS) provides the means for the remote transfer of in-vessel components and remote handling equipment between the Hot Cell Building and the Tokamak Building in ITER along pre-defined optimized trajectories. A first approach for CPRHS path optimization was previously proposed using line guidance as the navigation methodology to be adopted. This approach might not lead to feasible paths in new situations not considered during the previous work, as rescue operations. This paper addresses this problem by presenting a complementary approach for path optimization inspired in rigid body dynamics that takes full advantage of the rhombic like kinematics of the CPRHS. It also presents a methodology that maximizes the common parts of different trajectories in the same level of ITER buildings. The results gathered from 500 optimized trajectories are summarized. Conclusions and open issues are presented and discussed.