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ASME J of Nuclear Rad Sci. 2017;3(4):041001-041001-8. doi:10.1115/1.4036736.

Large sets of fluid temperature histories and a recently proposed thermal fatigue assessment procedure are employed in this paper to deliver more accurate statistics of predicted lives of pipes and their uncertainties under turbulent fluid mixing circumstances. The wide variety of synthetic fluid temperatures, generated with an improved spectral method, results in a set of estimated distributions of fatigue lives through linear one-dimensional (1D) heat diffusion, thermal stress estimates, and fatigue assessment codified rules. The results of the fatigue analysis indicate that, in order to avoid the inherent uncertainties due to comparatively short fluid temperature histories to the estimated fatigue lives, a conservative safe design against thermal fatigue could be attempted with the lower bounds of the predicted life distributions, such as the 5% probability life (5% of samples fail). The impact of the convection heat transfer coefficient on the predictions is also studied in a sensitivity analysis. This represents a detailed attempt to correlate the uncertainties in the physical fluid mixing conditions and heat transfer to the estimated fatigue life using spectral methods.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041002-041002-13. doi:10.1115/1.4036987.

Seawater was injected into the reactor cores following the accident at the Fukushima Daiichi nuclear power station. Saturated pool nucleate boiling heat transfer experiments with NaCl solution, natural seawater, and artificial seawater as well as distilled water were performed to examine the effects of salts on boiling heat transfer. The heat transfer surface was made of a printed copper circuit board. The boiling phenomena were recorded with a high-speed video camera. The surface-temperature distribution was measured with an infrared camera. In the experiments, the concentrations of the NaCl solutions and the artificial seawater were varied over a range of 3.5–10.0 wt. %. Boiling curves were well predicted with the Rohsenow correlation although large coalescent bubble formation was inhibited in the NaCl, natural seawater, and artificial seawater experiments. Deposits of calcium sulfate (CaSO4) on the heat transfer surface were observed in the experiments with artificial seawater. This formation of a deposit layer resulted in the initiation of a slow surface-temperature excursion at a heat flux lower than the usual critical heat flux (CHF). A unique relationship was confirmed between the salt concentrations of the artificial seawater in the bulk fluid and the vaporization rate at the surface at which the slow surface-temperature excursion initiated. This relationship suggested that if the bulk concentration of sea salts in the seawater exceeded 11 wt. %, the deposition of calcium sulfate on the heat transfer surface occurred even if the heat flux was zero.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041003-041003-9. doi:10.1115/1.4037118.
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This study characterized the magnitude, spatial profile, and frequency spectrum of thermal striping at a junction using a novel sodium-deployable optical fiber temperature sensor. Additionally, this study revealed for the first time the capability of performing cross correlation velocimetry (CCV) with an optical fiber to acquire fluid flow rates in a pipe. Optical fibers were encapsulated in stainless steel capillary tubes with an inert cover gas for high-temperature sodium deployment. Plots of temperature oscillation range as a function of two-dimensional space highlighted locations prone to mechanical failure for particular flow momentum ratios. The effect of inlet sodium temperature differential and bulk flow rate on thermal striping behavior was also explored. The power spectral density (PSD) revealed that the striping temperature oscillations occurred at frequencies ranging from 0.1 to 6 Hz. Finally, the bulk flow rate of liquid sodium was calculated from thermal striping's periodic temperature oscillations using cross correlation velocimetry for flow rates of 0.25–5.74 L/min.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041004-041004-11. doi:10.1115/1.4037115.

During normal operation in Canada deuterium uranium (CANDU®) reactors, the stress corrosion cracking (SCC) of fuel sheathing is mitigated effectively, in part, using a thin graphite-based coating known as CANDU lubricant (CANLUB). Mechanisms typically proposed for the demonstrated SCC mitigation offered by CANLUB include lubrication and/or chemical interactions. An additional possibility, that was recently suggested, involves the sequestering of iodine through its interaction with alkali metal and/or alkaline earth metal impurities in the CANLUB coating. This possibility is supported by the systematic analysis and testing in this paper, wherein three prevalent impurities (Na, Ca, and Mg) found in CANLUB were incorporated into SCC slotted ring experiments as metal oxides. When the amount of metal oxide (Na2O, CaO, or MgO) matched or exceeded the amount of iodine (6 mmol = 16 mg/cm3), Na2O and CaO protected the rings from corrosion whereas MgO enhanced their corrosion. When Zircaloy-4 sheathing is subjected to mechanical stress, high temperature, and high concentrations of iodine vapor, it is better protected by siloxane coatings than by graphite-CANLUB coatings. Consequently, since metal impurities (Na, Ca, and Mg) are found more abundantly in siloxane coatings than in graphite-CANLUB coatings, Zircaloy-4 slotted rings were coated with graphite-CANLUB containing Na, Ca, and/or Mg at those more abundant concentrations. Since these concentrations remain below 6 mmol, SCC test results suggest that the siloxane's superior adhesion is an essential first step in preventing corrosion induced by 6 mmol of iodine.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041005-041005-11. doi:10.1115/1.4036513.

A liquid Li jet flowing at 15 m/s under a high vacuum of 10−3 Pa is intended to serve as a beam target (Li target) in the planned International Fusion Materials Irradiation Facility (IFMIF). The engineering validation and engineering design activities (EVEDA) for the IFMIF are being implemented under the broader approach (BA) agreement. As a major activity of the Li target facility, the EVEDA Li test loop (ELTL) was constructed by the Japan Atomic Energy Agency. A stable Li target under the IFMIF conditions (Li temperature: 523.15 K, velocity: 15 m/s, and vacuum pressure: 10−3 Pa) was demonstrated using ELTL. This study focuses on a cavitationlike acoustic noise detected in a downstream conduit where the Li target flowed under vacuum conditions. This noise was investigated using acoustic-emission (AE) sensors installed at eight locations via acoustic wave guides. The sound intensity of the acoustic noise was examined against the cavitation number of the Li target. In addition, two types of frequency analysis, namely, fast Fourier transform (FFT) and continuous wavelet transform (CWT), were performed to characterize the acoustic noise. Owing to the acoustic noise's intermittency, high frequency, and the dependence on cavitation number, we conclude that this acoustic noise is generated when cavitation bubbles collapse and/or the structural material of the pipe is cracked because of the collapse of cavitation bubbles (cavitation pitting). The location of the cavitation was fundamental for presuming the mechanism. In this study, the propagation of acoustic waves among AE sensors placed at three locations was used to localize the cavitation and a method to determine the location of cavitation was formulated. As a result, we found that cavitation occurred only in a narrow area where the Li target impinged on the downstream conduit; therefore, we concluded that this cavitation was induced by the impingement. The design of the downstream conduit of the IFMIF Li target facility should be tackled in future based on information obtained in this study.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041006-041006-10. doi:10.1115/1.4037189.

A finite element method (FEM) is applied to investigate the thermal conductivity of polycrystalline UO2. The influences of microstructure are especially important for UO2 due to the severe structural changes under irradiation conditions. In this study, we have investigated the influences of microstructures on the thermal conductivity of polycrystalline UO2 using FEM. The temperature profile of fuel pellet with different microstructures during service is also investigated. The thermal conductivity increases with increasing grain size. The grain size distribution has obvious influence on the thermal conductivity especially when there are pores in the polycrystal. The influences of porosity and pore size are very sensitive to the position of the pores. The results obtained in this study are useful for the prediction of property changes of UO2 fuel in pile and important to gain some design guidance to tune the properties through the control of the microstructure.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041007-041007-11. doi:10.1115/1.4037095.

A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) varying gas gaps. The experiment was irradiated at the advanced test reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program, which supports the development of the very high-temperature gas-cooled reactor under the advanced reactor technologies project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) tristructural-isotropic-fueled compacts were inserted into 12 separate capsules for the experiment. The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using PIE-measured time (fast neutron fluence) varying gas gaps and compare with experimentally measured thermocouple (TC) data. PIE-measured experimental data were used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all the fuel compacts, graphite holders, and all of the graphite rings used. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each nonfuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for each capsule.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041008-041008-8. doi:10.1115/1.4037204.

Nondestructive testing (NDT) techniques are widely used as a reliable way for preventing failures and helping in the maintenance design and operation of critical infrastructures and complex industrial plants as nuclear power plants (NPPs). Among the NDT techniques, guided waves (GWs) are a very promising technology for such applications. GWs are structure-borne ultrasonic waves propagating along the structure confined and guided by its geometric boundaries. Testing using GWs is able to find defect locations through long-range screening using low-frequency waves (from 5 to 250 kHz). The technology is regularly used for pipe testing in the oil and gas industry. In the nuclear industry, regulators are working to standardize monitoring and inspection procedures. To use the technology inside an active plant, operators must solve issues like high temperatures (up to more than 300 °C inside a light-water reactor's primary piping), high wall thickness of components in the primary circuit, and characteristic defect typologies. Magnetostrictive sensors are expected to overcome such issues due to their physical properties, namely, robust constitution and simplicity. Recent experimental results have demonstrated that magnetostrictive transducers can withstand temperatures close to 300 °C. In this paper, the GW technology will be introduced in the context of NPPs. Some experimental tests conducted using such a methodology for steel pipe having a complex structure will be described, and open issues related to high-temperature guided wave applications (e.g., wave velocity or amplitude fluctuations during propagation in variable temperature components) will be discussed.

Topics: Waves , Pipes , Steel
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041009-041009-13. doi:10.1115/1.4037188.

The initial design of ITER incorporated the use of carbon fiber composites in high heat flux regions and tungsten was used for low heat flux regions. The current design includes tungsten for both these regions. The present work includes thermal hydraulic modeling and analysis of ex-vessel loss of coolant accident (LOCA) for the divertor (DIV) cooling system. The purpose of this study is to show that the new concept of full tungsten divertor is able to withstand in the accident scenarios. The code used in this study is RELAP/SCADAPSIM/MOD 4.0. A parametric study is also carried out with different in-vessel break sizes and ex-vessel break locations. The analysis discusses a number of safety concerns that may result from the accident scenarios. These concerns include vacuum vessel (VV) pressurization, divertor temperature profile, passive decay heat removal capability of structure, and pressurization of tokamak cooling water system. The results show that the pressures and temperatures are kept below design limits prescribed by ITER organization.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041010-041010-6. doi:10.1115/1.4037163.

Cementitious materials for the construction of a geological repository of radioactive waste alter the pH of groundwater to a highly alkaline condition (pH ≈ 13). While this alkaline groundwater dissolves silicate minerals, the soluble silicic acid polymerizes or deposits on the surface of rock with the decrease in pH by mixing with the surrounding groundwater (pH = 8). In particular, the deposition of silicic acid leads to a clogging effect in flow-paths, which retards the migration of radionuclides. This study estimated the clogging of silicic acid in flow-paths with the one-dimensional advection–dispersion model considering the deposition rate constants evaluated in our previous study. As some of the most important parameters, these estimations focused on the initial supersaturated concentration of silicic acid and the density of deposited minerals. As a result, the aperture of flow-paths (initial width: 0.1 mm, flow-rate: 5 m/y, initial supersaturated concentration of silicic acid: 0.01, 0.1 and 1.0 mM) was almost clogged within about 200 y by the deposition of silicic acid. The period for the clogging became shorter under the conditions of higher initial supersaturated concentration and lower density of deposited minerals. In other words, the use of cementitious materials for constructing the repository might produce a retardation effect of radionuclide migration by the deposition/clogging processes of the supersaturated silicic acid.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041011-041011-9. doi:10.1115/1.4036892.

To respond to the urgent needs of verification, training, and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities, and research institutes, the FSSAMG verification and drill system is developed. The FSSAMG includes comprehensive scenarios under power condition, shutdown condition, spent fuel pool (SFP) condition, and refueling conditions. This article summarized the research and development of validation and drill system for FSSAMG by using the severe accident analysis computer code modular accident analysis program 5 (MAAP5). Realistic accident scenarios can be verified and exercised in the developed system to support FSSAMG training, drill, examination, and verification.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041012-041012-8. doi:10.1115/1.4037031.

Fukushima accident has raised a strong concern and apprehension about the safety of a nuclear reactor failing to remove the decay heat following an extreme event. After Fukushima accident, the reactor designers worldwide analyzed the safety margin of the existing and new generation nuclear power plants for such an event. Advanced heavy water reactor (AHWR), designed in India, was also analyzed for even more severe conditions than occurred at Fukushima. AHWR equipped with several passive systems showed its robustness against this type of scenarios. However, several new passive systems were incorporated in AHWR design for maintaining the integrity of the reactor at least for 7 days as a grace period. A passive moderator cooling system (PMCS) and a passive endshield cooling system (PECS) were among the newly introduced safety system in AHWR. An experimental test facility simulating the prolonged station blackout (SBO) case in AHWR has been designed and built. Experiments have been performed in the test facility for simulated conditions of prolonged SBO. The current study shows the performance of AHWR during prolonged SBO case through simulation in the integral test facility. The results indicate that AHWR design is capable of removing decay heat for prolonged period without operator interference.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041013-041013-8. doi:10.1115/1.4036985.

In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured temperatures of RPV and the air outside of biological shielding concrete. The temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041014-041014-10. doi:10.1115/1.4037094.

For the purpose of nuclear safety analysis, a reactive flow solver has been developed to determine the hazardous potential of large-scale hydrogen explosions. Without using empirical transition criteria, the whole combustion process including deflagration-to-detonation transition (DDT) is computed within a single solver framework. In this paper, we present massively parallelized three-dimensional explosion simulations in a full-scale pressurized water reactor (PWR) of the Konvoi type. Several generic DDT scenarios in globally lean hydrogen–air mixtures are examined to assess the importance of different input parameters. It is demonstrated that the explosion process is highly sensitive to mixture composition, ignition location, and thermodynamic initial conditions. Pressure loads on the confining structure show a profoundly dynamic behavior depending on the position in the containment. Computational cost can effectively be reduced through adaptive mesh refinement (AMR).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041015-041015-11. doi:10.1115/1.4036984.

The presented work aims to improve computational fluid dynamics (CFD) explosion modeling for lean hydrogen–air mixtures on under-resolved grids. Validation data are obtained from an entirely closed laboratory-scale explosion channel (GraVent facility). Investigated hydrogen–air concentrations range from 6 to 19 vol %. Initial conditions are p = 0.1 MPa and T = 293 K. Two highly time-resolved optical measurement techniques are applied simultaneously: (1) 10 kHz shadowgraphy captures line-of-sight integrated macroscopic flame propagation and (2) 20 kHz planar laser-induced fluorescence of the OH radical (OH-PLIF) resolves microscopic flame topology without line-of-sight integration. This paper presents the experiment, measurement techniques, data evaluation methods, and simulation results. The evaluation methods encompass the determination of flame tip velocity over distance and a detailed time-resolved quantification of the flame topology based on OH-PLIF images. One parameter is the length of wrinkled flame fronts in the OH-PLIF plane obtained through automated postprocessing. It reveals the expected enlargement of flame surface area by instabilities on a microscopic level. A strong effect of mixture composition is observed. Simulations based on the new model formulation, incorporating the microscopic enlargement of the flame front, show a promising behavior, where the impact of the augmented flame front on the observed flame front velocities can be detected.

Topics: Flames , Hydrogen , Simulation
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041016-041016-9. doi:10.1115/1.4036737.

An important requirement for Generation IV Nuclear Power Plant (NPP) design is the control system, which enables part power operability. The choices of control system methods must ensure variation of load without severe drawbacks on cycle performance. The objective of this study is to assess the control of the NPP under part power operations. The cycles of interest are the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). Control strategies are proposed for NPPs but the focus is on the strategies that result in part power operation using the inventory control method. First, results explaining the performance and load limiting factors of the inventory control method are documented; subsequently, the transient part power performances are also documented. The load versus efficiency curves were also derived from varying the load to understand the efficiency penalties. This is carried out using a modeling and performance simulation tool designed for this study. Results show that the ICR takes ∼102% longer than the SCR to reduce the load to 50% in design point (DP) performance conditions for similar valve flows, which correlates with the volumetric increase for the ICR inventory tank. The efficiency penalties are comparable for both cycles at 50% part power, whereby a 22% drop in cycle efficiency was observed and indicates limiting time at very low part power. The analyses intend to aid the development of cycles for Generation IV NPPs specifically gas cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs), where helium is the coolant.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041017-041017-8. doi:10.1115/1.4036983.

The control system for generation IV nuclear power plant (NPP) design must ensure load variation when changes to critical parameters affect grid demand, plant efficiency, and component integrity. The objective of this study is to assess the load following capabilities of cycles when inventory pressure control is utilized. Cycles of interest are simple cycle recuperated (SCR), intercooled cycle recuperated (ICR), and intercooled cycle without recuperation (IC). First, part power performance of the IC is compared to results of the SCR and ICR. Subsequently, the load following capabilities are assessed when the cycle inlet temperatures are varied. This was carried out using a tool designed for this study. Results show that the IC takes ∼2.7% longer than the ICR to reduce the power output to 50% when operating in design point (DP) for similar valve flows, which correlates to the volumetric increase for the IC inventory storage tank. However, the ability of the IC to match the ICR's load following capabilities is severely hindered because the IC is most susceptible to temperature variation. Furthermore, the IC takes longer than the SCR and ICR to regulate the reactor power by a factor of 51 but this is severely reduced, when regulating NPP power output. However, the IC is the only cycle that does not compromise reactor integrity and cycle efficiency when regulating the power. The analyses intend to aid the development of cycles specifically gas-cooled fast reactors (GFRs) and very high temperature reactors (VHTRs), where helium is the coolant.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(4):041018-041018-7. doi:10.1115/1.4037262.

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2017;3(4):044501-044501-3. doi:10.1115/1.4037078.

The decommissioning of nuclear facilities in Indonesia are mainly based on Nuclear Energy Act of 1997 on, Nuclear Energy Regulatory Agency (BAPETEN) Chairman Regulation (BCR) No. 4 of 2009, and Government Regulation (GR) No. 2 of 2014 that cover general provisions and licensing requirements of decommissioning and the detailed requirements and guidelines for preparing the decommissioning plan and licensee applications. BCR No. 4 of 2009 was developed based on the adoption and adaption from the International Atomic Energy Agency (IAEA) Safety Guide, WS-G-2.1. Currently, one of three research reactors, Training Research Isotopes General Atomics (TRIGA) research reactor 2 MW (the oldest which went critical at 250 kW in 1964, and was operated at maximum in 1971, and was upgraded to 2 MW in 2000), has operated for 45 y, but there is no decision for decommissioning this reactor yet. Indonesia has experience in decommissioning of the phosphoric acid purification facility of the Gresik petrochemical plant. Some aspects of decommissioning, which have been successfully addressed to date, are: regulation, communication, and decommissioning team. Development of human resources, technological capability, and information flow from more advanced countries are an important factor for the future of the nuclear facility decommissioning plan in Indonesia. Some regulations have still not anticipated all the regulatory challenges that might be encountered in the near future. More regulations and guidance are needed to be established by BAPETEN in order to complete the current regulations so that problems can all be anticipated.

Commentary by Dr. Valentin Fuster

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